PRA: an evaluation of state-of-the-art

International Nuclear Information System (INIS)

Joksimovich, V.

1985-01-01

Some elements of the probabilistic risk assessment (PRA) methodology can be characterized as mature and are even ready for some kind of a standardization effort. Other elements are still, however, in a rapid state of evolution. Questions are continuously being asked regarding maturity of PRA techniques vis-a-vis a regulatory decision-making process. Establishing a framework for evaluating state-of-the-art in any technological field is a challenging task. An implementation of a selected framework to a satisfactory conclusion is a monumental task. Of course, these types of issues can be discussed meaningfully only if they are tied to a particular application. The author's participation in the NSF-sponsored risk assessment project is discussed in the paper. The evaluation employed here makes use of the following five evaluation criteria: logical soundness, completeness, accuracy, acceptability, and practicality

Probabilistic risk assessment (PRA) update in light of the accident at Fukushima Daiichi Nuclear Power Station - 15461

International Nuclear Information System (INIS)

Maeda, K.; Abe, H.; Hirokawa, N.; Satou, C.

2015-01-01

We have performed internal and external event probabilistic risk assessments (PRA) for boiling water reactor power nuclear plants to identify the important accident sequence groups and to evaluate the effectiveness of the additional severe accident measures, regarding to the new regulatory requirements implemented after the accident at Fukushima Daiichi Nuclear Power Station in Japan in 2011. In addition, we will further update our PRA by extracting problems and improvements from the current PRA, by catching up the state-of-the-art knowledge, modern PRA methodologies in order to contribute voluntarily to safety improvement as well as to comply with regulations. In this document, prior to the extensive PRA updates, we would describe technical contents and qualitative results about PRA updates that have been performed preliminary so far, especially about the external event (seismic) PRA and how to model the additionally deployed severe accident measures (e.g. power supply car, fire engine) so that they can be function external hazards, such as component failure rate of equipment, human reliability 'out of control room', and mission time extension. (authors)

Review of PRA methodology for LMFBR

International Nuclear Information System (INIS)

Yang, J. E.

1999-02-01

Probabilistic Risk Assessment (PRA) has been widely used as a tool to evaluate the safety of NPPs (Nuclear Power Plants), which are in the design stage as well as in operation. Recently, PRA becomes one of the licensing requirements for many existing and new NPPs. KALIMER is a Liquid Metal Fast Breeder Reactor (LMFBR) being developed by KAERI. Since the design concept of KALIMER is similar to that of the PRISM plant developed by GE, it would be appropriate to review the PRA methodology of PRISM as the first step of KALIMER PRA. Hence, in this report summarizes the PRA methodology of PRISM plant, and the required works for the PSA of KALIMER based on the reviewed results. The PRA technology of PRISM plant consists of following five major tasks: (1) development of initiating event list, (2) development of system event tree, (3) development of core response event tree, (4) development of containment response event tree, and (5) consequences and risk estimation. The estimated individual and societal risk measures show that the risk from a PRISM module is substantially less than the NRC goal. Each task is compared to the PRA methodology of Light Water Reactor (LWR)/Pressurized Heavy Water Reactor (PHWR). In the report, each task of PRISM PRA methodology is reviewed and compared to the corresponding part of LWR/PHWR PSA performed in Korea. The parts that are not modeled appropriately in PRISM PRA are identified, and the recommendations for KALIMER PRA are stated. (author). 14 refs., 9 tabs., 4 figs

Linkage of PRA models. Phase 1, Results

Energy Technology Data Exchange (ETDEWEB)

Smith, C.L.; Knudsen, J.K.; Kelly, D.L.

1995-12-01

The goal of the Phase I work of the ``Linkage of PRA Models`` project was to postulate methods of providing guidance for US Nuclear Regulator Commission (NRC) personnel on the selection and usage of probabilistic risk assessment (PRA) models that are best suited to the analysis they are performing. In particular, methods and associated features are provided for (a) the selection of an appropriate PRA model for a particular analysis, (b) complementary evaluation tools for the analysis, and (c) a PRA model cross-referencing method. As part of this work, three areas adjoining ``linking`` analyses to PRA models were investigated: (a) the PRA models that are currently available, (b) the various types of analyses that are performed within the NRC, and (c) the difficulty in trying to provide a ``generic`` classification scheme to groups plants based upon a particular plant attribute.

Linkage of PRA models. Phase 1, Results

International Nuclear Information System (INIS)

Smith, C.L.; Knudsen, J.K.; Kelly, D.L.

1995-12-01

The goal of the Phase I work of the ''Linkage of PRA Models'' project was to postulate methods of providing guidance for US Nuclear Regulator Commission (NRC) personnel on the selection and usage of probabilistic risk assessment (PRA) models that are best suited to the analysis they are performing. In particular, methods and associated features are provided for (a) the selection of an appropriate PRA model for a particular analysis, (b) complementary evaluation tools for the analysis, and (c) a PRA model cross-referencing method. As part of this work, three areas adjoining ''linking'' analyses to PRA models were investigated: (a) the PRA models that are currently available, (b) the various types of analyses that are performed within the NRC, and (c) the difficulty in trying to provide a ''generic'' classification scheme to groups plants based upon a particular plant attribute

PRA: A Perspective on Strengths, Current Limitations, And Possible Improvements

International Nuclear Information System (INIS)

Mosleh, Ail

2014-01-01

Probabilistic risk assessment (PRA) has been used in various technological fields to assist regulatory agencies, managerial decision makers, and systems designers in assessing and mitigating the risks inherent in these complex arrangements. Has PRA delivered on its promise? How do we gage PRA performance? Are our expectations about value of PRA realistic? Are there disparities between what we get and what we think we are getting form PRA and its various derivatives? Do current PRAs reflect the knowledge gained from actual events? How do we address potential gaps? These are some of the questions that have been raised over the years since the inception of the field more than forty years ago. This paper offers a brief assessment of PRA as a technical discipline in theory and practice, its key strengths and weaknesses, and suggestions on ways to address real and perceived shortcomings

The tsunami probabilistic risk assessment (PRA). Example of accident sequence analysis of tsunami PRA according to the standard for procedure of tsunami PRA for nuclear power plants

International Nuclear Information System (INIS)

Ohara, Norihiro; Hasegawa, Keiko; Kuroiwa, Katsuya

2013-01-01

After the Fukushima Daiichi nuclear power plant (NPP) accident, standard for procedure of tsunami PRA for NPP had been established by the Standardization Committee of AESJ. Industry group had been conducting analysis of Tsunami PRA for PWR based on the standard under the cooperation with electric utilities. This article introduced overview of the standard and examples of accident sequence analysis of Tsunami PRA studied by the industry group according to the standard. The standard consisted of (1) investigation of NPP's composition, characteristics and site information, (2) selection of relevant components for Tsunami PRA and initiating events and identification of accident sequence, (3) evaluation of Tsunami hazards, (4) fragility evaluation of building and components and (5) evaluation of accident sequence. Based on the evaluation, countermeasures for further improvement of safety against Tsunami could be identified by the sensitivity analysis. (T. Tanaka)

Management and Organization Influences in PRA

International Nuclear Information System (INIS)

Gertman, D.I.; Hallbert, B. P.; Blackman, H. S.

1998-01-01

The authors present a research program which aimed at increasing the quality of comprehensiveness of contemporary PRA (Probability Risk Assessment) by providing a tool that allows for incorporating M and O in PRA, at improving the quality of NRC assessments, at conducting research to support the risk informed regulation process, at identifying impact of management and organization, safety culture, workplace environment, down-sizing and deregulation on human performance and reliability

PRA: A PERSPECTIVE ON STRENGTHS, CURRENT LIMITATIONS, AND POSSIBLE IMPROVEMENTS

Directory of Open Access Journals (Sweden)

ALI MOSLEH

2014-02-01

Full Text Available Probabilistic risk assessment (PRA has been used in various technological fields to assist regulatory agencies, managerial decision makers, and systems designers in assessing and mitigating the risks inherent in these complex arrangements. Has PRA delivered on its promise? How do we gage PRA performance? Are our expectations about value of PRA realistic? Are there disparities between what we get and what we think we are getting form PRA and its various derivatives? Do current PRAs reflect the knowledge gained from actual events? How do we address potential gaps? These are some of the questions that have been raised over the years since the inception of the field more than forty years ago. This paper offers a brief assessment of PRA as a technical discipline in theory and practice, its key strengths and weaknesses, and suggestions on ways to address real and perceived shortcomings.

Standardized procedure for tsunami PRA by AESJ

International Nuclear Information System (INIS)

Kirimoto, Yukihiro; Yamaguchi, Akira; Ebisawa, Katsumi

2013-01-01

After Fukushima Accident (March 11, 2011), the Atomic Energy Society of Japan (AESJ) started to develop the standard of Tsunami Probabilistic Risk Assessment (PRA) for nuclear power plants in May 2011. As Japan is one of the countries with frequent earthquakes, a great deal of efforts has been made in the field of seismic research since the early stage. To our regret, the PRA procedures guide for tsunami has not yet been developed although the importance is held in mind of the PRA community. Accordingly, AESJ established a standard to specify the standardized procedure for tsunami PRA considering the results of investigation into the concept, the requirements that should have and the concrete methods regarding tsunami PRA referring the opinions of experts in the associated fields in December 2011 (AESJ-SC-RK004:2011). (author)

Results of the Level 1 probabilistic risk assessment (PRA) of internal events for heavy water production reactors

International Nuclear Information System (INIS)

Tinnes, S.P.; Cramer, D.S.; Logan, V.E.; Topp, S.V.; Smith, J.A.; Brandyberry, M.D.

1990-01-01

A full-scope probabilistic risk assessment (PRA) is being performed for the Savannah River site (SRS) production reactors. The Level 1 PRA for the K Reactor has been completed and includes the assessment of reactor systems response to accidents and estimates of the severe core melt frequency (SCMF). The internal events spectrum includes those events related directly to plant systems and safety functions for which transients or failures may initiate an accident. The SRS PRA has three principal objectives: improved understanding of SRS reactor safety issues through discovery and understanding of the mechanisms involved. Improved risk management capability through tools for assessing the safety impact of both current standard operations and proposed revisions. A quantitative measure of the risks posed by SRS reactor operation to employees and the general public, to allow comparison with declared goals and other societal risks

Probabilistic risk assessment course documentation. Volume 2. Probability and statistics for PRA applications

International Nuclear Information System (INIS)

Iman, R.L.; Prairie, R.R.; Cramond, W.R.

1985-08-01

This course is intended to provide the necessary probabilistic and statistical skills to perform a PRA. Fundamental background information is reviewed, but the principal purpose is to address specific techniques used in PRAs and to illustrate them with applications. Specific examples and problems are presented for most of the topics

PRA and Risk Informed Analysis

International Nuclear Information System (INIS)

Bernsen, Sidney A.; Simonen, Fredric A.; Balkey, Kenneth R.

2006-01-01

The Boiler and Pressure Vessel Code (BPVC) of the American Society of Mechanical Engineers (ASME) has introduced a risk based approach into Section XI that covers Rules for Inservice Inspection of Nuclear Power Plant Components. The risk based approach requires application of the probabilistic risk assessments (PRA). Because no industry consensus standard existed for PRAs, ASME has developed a standard to evaluate the quality level of an available PRA needed to support a given risk based application. The paper describes the PRA standard, Section XI application of PRAs, and plans for broader applications of PRAs to other ASME nuclear codes and standards. The paper addresses several specific topics of interest to Section XI. Important consideration are special methods (surrogate components) used to overcome the lack of PRA treatments of passive components in PRAs. The approach allows calculations of conditional core damage probabilities both for component failures that cause initiating events and failures in standby systems that decrease the availability of these systems. The paper relates the explicit risk based methods of the new Section XI code cases to the implicit consideration of risk used in the development of Section XI. Other topics include the needed interactions of ISI engineers, plant operating staff, PRA specialists, and members of expert panels that review the risk based programs

PRISIM: a computer program that makes PRA useful

International Nuclear Information System (INIS)

Fussell, J.B.; Campbell, D.J.; Glynn, J.C.; Burdick, G.R.

1986-01-01

PRISIM is an IBM personal computer program that translates probabilistic risk assessment (PRA) information and calculates additional PRA type information for use by those who are not PRA experts. Specifically, PRISIM was developed for the US Nuclear Regulatory Commission for use by their resident inspectors at nuclear power plants. Inspector activities are either scheduled or are in response to a particular status of a plant. PRISIM is useful for either activity

76 FR 81998 - Methodology for Low Power/Shutdown Fire PRA

Science.gov (United States)

2011-12-29

... NUCLEAR REGULATORY COMMISSION [NRC-2011-0295] Methodology for Low Power/Shutdown Fire PRA AGENCY..., ``Methodology for Low Power/Shutdown Fire PRA--Draft Report for Comment.'' DATES: Submit comments by March 01... risk assessment (PRA) method for quantitatively analyzing fire risk in commercial nuclear power plants...

Uses of PRA in nuclear reactor regulation

International Nuclear Information System (INIS)

Congel, F.

1987-01-01

For the past five years, more than ten probabilistic risk assessment (PRA) studies were conducted by the owners of nuclear utilities and were submitted for the review of US Nuclear Regulatory Commission staff. These PRA studies were reviewed under various types of regulatory activities depending on the nature of plant licensing stage. The reviews of these PRAs provided very valuable uses to both the staff and the licensees on safety matters of the plant operation. The licensees developed perspectives using PRA models on the safety profiles of their plants. These PRA perspectives influenced licensees' major decisions to implement improvements to plant design and operating and emergency procedures to reduce and/or eliminate the plant's vulnerability to core damage accidents. The staff's review of these PRAs particularly emphasized the dominant accident sequences. The resulting findings led to the identification of dominant risk contributors, critical areas of plant locations, mechanisms leading to potential early containment failures, and instances of noncompliances of staff's deterministic criteria. Specific examples include single failure criterion and separation requirements to assess the need for any additional measures to further improve the safety of the plant. Some of these PRAs were reviewed under regulatory activities other than safety review such as environmental review, final design review, and licensing hearings. Most importantly, the risk profiles of generic PRAs will continue to be used in reviewing and evaluating unresolved safety issues and other generic issues. The major regulatory uses of PRAs, a summary of full scope PRA review, a summary of plant improvements as a result of PRA reviews, and the future role of PRA reviews are presented

77 FR 10576 - Methodology for Low Power/Shutdown Fire PRA

Science.gov (United States)

2012-02-22

... NUCLEAR REGULATORY COMMISSION [NRC-2011-0295] Methodology for Low Power/Shutdown Fire PRA AGENCY.../Shutdown Fire PRA.'' In response to request from members of the public, the NRC is extending the public... risk assessment (PRA) method for quantitatively analyzing fire risk in commercial nuclear power plants...

Results of the Level 1 probabilistic risk assessment (PRA) of internal events for heavy water production reactors (U)

International Nuclear Information System (INIS)

Tinnes, S.P.; Cramer, D.S.; Logan, V.E.; Topp, S.V.; Smith, J.A.; Brandyberry, M.D.

1990-01-01

This paper reports on a full-scope probabilistic risk assessment (PRA) performed for the Savannah River Site (SRS) production reactors. The Level 1 PRA for the K Reactor has been completed and includes the assessment of reactor systems response to accidents and estimates of the severe core melt frequency (SCMF). The internal events spectrum includes those events related directly to plant systems and safety functions for which transients or failures may initiate an accident

SHARP - a framework for incorporating human interactions into PRA studies

International Nuclear Information System (INIS)

Hannaman, G.W.; Joksimovich, V.; Spurgin, A.J.; Worledge, D.H.

1985-01-01

Recently, increased attention has been given to understanding the role of humans in the safe operation of nuclear power plants. By virtue of the ability to combine equipment reliability with human reliability probabilistic risk assessment (PRA) technology was deemed capable of providing significant insights about the contributions of human interations in accident scenarios. EPRI recognized the need to strengthen the methodology for incorporating human interactions into PRAs as one element of their broad research program to improve the credibility of PRAs. This research project lead to the development and detailed description of SHARP (Systematic Human Application Reliability Procedure) in EPRI NP-3583. The objective of this paper is to illustrate the SHARP framework. This should help PRA analysts state more clearly their assumptions and approach no matter which human reliability assessment technique is used. SHARP includes a structure of seven analysis steps which can be formally or informally performed during PRAs. The seven steps are termed definition, screening, breakdown, representation, impact assessment, quantification, and documentation

Application of multivariate statistical technique for hydrogeochemical assessment of groundwater within the Lower Pra Basin, Ghana

Science.gov (United States)

Tay, C. K.; Hayford, E. K.; Hodgson, I. O. A.

2017-06-01

Multivariate statistical technique and hydrogeochemical approach were employed for groundwater assessment within the Lower Pra Basin. The main objective was to delineate the main processes that are responsible for the water chemistry and pollution of groundwater within the basin. Fifty-four (54) (No) boreholes were sampled in January 2012 for quality assessment. PCA using Varimax with Kaiser Normalization method of extraction for both rotated space and component matrix have been applied to the data. Results show that Spearman's correlation matrix of major ions revealed expected process-based relationships derived mainly from the geochemical processes, such as ion-exchange and silicate/aluminosilicate weathering within the aquifer. Three main principal components influence the water chemistry and pollution of groundwater within the basin. The three principal components have accounted for approximately 79% of the total variance in the hydrochemical data. Component 1 delineates the main natural processes (water-soil-rock interactions) through which groundwater within the basin acquires its chemical characteristics, Component 2 delineates the incongruent dissolution of silicate/aluminosilicates, while Component 3 delineates the prevalence of pollution principally from agricultural input as well as trace metal mobilization in groundwater within the basin. The loadings and score plots of the first two PCs show grouping pattern which indicates the strength of the mutual relation among the hydrochemical variables. In terms of proper management and development of groundwater within the basin, communities, where intense agriculture is taking place, should be monitored and protected from agricultural activities. especially where inorganic fertilizers are used by creating buffer zones. Monitoring of the water quality especially the water pH is recommended to ensure the acid neutralizing potential of groundwater within the basin thereby, curtailing further trace metal

Loss of coolant accident (LOCA) analysis for McMaster Nuclear Reactor through probabilistic risk assessment (PRA)

Energy Technology Data Exchange (ETDEWEB)

Ha, T.; Garland, W.J. [McMaster Univ., Dept. of Engineering Physics, Hamilton, Ontario (Canada)]. E-mail: hats@mcmaster.ca

2006-07-01

A probabilistic risk assessment (PRA) was conducted for the loss of coolant accident (LOCA) sequence in the McMaster Nuclear Reactor (MNR). A level 1 PRA was completed including event sequence modeling, system modeling, and quantification. To support the quantification of the accident sequence identified, data analysis using the Bayesian method and human reliability analysis (HRA) using the ASEP approach were performed. Since human performance in research reactors is significantly different from that in power reactors, a different time-oriented HRA model was proposed and applied for the estimation of the human error probability (HEP) of core relocation. This HEP estimate was less than that by the ASEP approach by a factor of about 2. These two HEP estimates were used for sensitivity analysis, and modeling uncertainty in the PRA models was quantified. This showed the necessity of appropriate human reliability models in PRA for research reactors. This method could be implemented for the operators' actions which require extensive manual execution with little cognitive load, as might be the case for some maintenance operations in power reactors. (author)

PRA-Code Upgrade to Handle a Generic Problem

International Nuclear Information System (INIS)

Wilson, J. R.

1999-01-01

During the probabilistic risk assessment (PRA) for the proposed Yucca Mountain nuclear waste repository, a problem came up that could not be handled by most PRA computer codes. This problem deals with dependencies between sequential events in time. Two similar scenarios that illustrate this problem are LOOP nonrecovery and sequential wearout failures with units of time. The purpose of this paper is twofold: To explain the problem generically, and to show how the PRA code at the INEEL, SAPHIRE, has been modified to solve this problem correctly

Seabrook Station Level 2 PRA Update to Include Accident Management

International Nuclear Information System (INIS)

Lutz, Robert; Lucci, Melissa; Kiper, Kenneth; Henry, Robert

2006-01-01

A ground-breaking study was recently completed as part of the Seabrook Level 2 PRA update. This study updates the post-core damage phenomena to be consistent with the most recent information and includes accident management activities that should be modeled in the Level 2 PRA. Overall, the result is a Level 2 PRA that fully meets the requirements of the ASME PRA Standard with respect to modeling accident management in the LERF assessment and NRC requirements in Regulatory Guide 1.174 for considering late containment failures. This technical paper deals only with the incorporation of operator actions into the Level 2 PRA based on a comprehensive study of the Seabrook Station accident response procedures and guidance. The paper describes the process used to identify the key operator actions that can influence the Level 2 PRA results and the development of success criteria for these key operator actions. This addresses a key requirement of the ASME PRA Standard for considering SAMG. An important benefit of this assessment was the identification of Seabrook specific accident management insights that can be fed back into the Seabrook Station accident management procedures and guidance or the training provided to plant personnel for these procedures and guidance. (authors)

The Evaluation of the Adequacy of PRA Results for Risk-informed Decision Makings With Respect to Incompleteness

International Nuclear Information System (INIS)

Kang, Kyungmin; Jae, Moosung

2007-01-01

PRA(Probabilistic Risk Assessment), as a quantitative tool, has many strengths as well as weaknesses. There are several limitations on the use of PRA techniques for risk modeling and analysis. First, the true values of most model inputs are unknown. Ideally, probability distribution models are well developed and assigned to the unknown input parameters to reflect the analyst's state of knowledge of the values of this input parameter. The problem of overconfidence and lack of confidence in the values of certain model input parameters can lead to inaccurate PRA results. Secondly, the analyst's lack of knowledge of a system's practical application as opposed to its theoretical operation can lead to modeling errors. The quality of PRAs has been addressed by a number of regulatory and industry organizations Some have argued that a good PRA should be a complete, full scope, three level PRA, while others have claimed that the quality of a PRA should be measured with respect to the application and decision supported. we show by way of an example that the adequacy of a PRA results is important to risk-informed decision making process and should be measured with respect to the application and decision supported

Advanced Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) Technical Exchange Meeting

Energy Technology Data Exchange (ETDEWEB)

Smith, Curtis [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2013-09-01

During FY13, the INL developed an advanced SMR PRA framework which has been described in the report Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) Detailed Technical Framework Specification, INL/EXT-13-28974 (April 2013). In this framework, the various areas are considered: Probabilistic models to provide information specific to advanced SMRs Representation of specific SMR design issues such as having co-located modules and passive safety features Use of modern open-source and readily available analysis methods Internal and external events resulting in impacts to safety All-hazards considerations Methods to support the identification of design vulnerabilities Mechanistic and probabilistic data needs to support modeling and tools In order to describe this framework more fully and obtain feedback on the proposed approaches, the INL hosted a technical exchange meeting during August 2013. This report describes the outcomes of that meeting.

Spatially Informed Plant PRA Models for Security Assessment

International Nuclear Information System (INIS)

Wheeler, Timothy A.; Thomas, Willard; Thornsbury, Eric

2006-01-01

Traditional risk models can be adapted to evaluate plant response for situations where plant systems and structures are intentionally damaged, such as from sabotage or terrorism. This paper describes a process by which traditional risk models can be spatially informed to analyze the effects of compound and widespread harsh environments through the use of 'damage footprints'. A 'damage footprint' is a spatial map of regions of the plant (zones) where equipment could be physically destroyed or disabled as a direct consequence of an intentional act. The use of 'damage footprints' requires that the basic events from the traditional probabilistic risk assessment (PRA) be spatially transformed so that the failure of individual components can be linked to the destruction of or damage to specific spatial zones within the plant. Given the nature of intentional acts, extensive modifications must be made to the risk models to account for the special nature of the 'initiating events' associated with deliberate adversary actions. Intentional acts might produce harsh environments that in turn could subject components and structures to one or more insults, such as structural, fire, flood, and/or vibration and shock damage. Furthermore, the potential for widespread damage from some of these insults requires an approach that addresses the impacts of these potentially severe insults even when they occur in locations distant from the actual physical location of a component or structure modeled in the traditional PRA. (authors)

Medical Updates Number 5 to the International Space Station Probability Risk Assessment (PRA) Model Using the Integrated Medical Model

Science.gov (United States)

Butler, Doug; Bauman, David; Johnson-Throop, Kathy

2011-01-01

The Integrated Medical Model (IMM) Project has been developing a probabilistic risk assessment tool, the IMM, to help evaluate in-flight crew health needs and impacts to the mission due to medical events. This package is a follow-up to a data package provided in June 2009. The IMM currently represents 83 medical conditions and associated ISS resources required to mitigate medical events. IMM end state forecasts relevant to the ISS PRA model include evacuation (EVAC) and loss of crew life (LOCL). The current version of the IMM provides the basis for the operational version of IMM expected in the January 2011 timeframe. The objectives of this data package are: 1. To provide a preliminary understanding of medical risk data used to update the ISS PRA Model. The IMM has had limited validation and an initial characterization of maturity has been completed using NASA STD 7009 Standard for Models and Simulation. The IMM has been internally validated by IMM personnel but has not been validated by an independent body external to the IMM Project. 2. To support a continued dialogue between the ISS PRA and IMM teams. To ensure accurate data interpretation, and that IMM output format and content meets the needs of the ISS Risk Management Office and ISS PRA Model, periodic discussions are anticipated between the risk teams. 3. To help assess the differences between the current ISS PRA and IMM medical risk forecasts of EVAC and LOCL. Follow-on activities are anticipated based on the differences between the current ISS PRA medical risk data and the latest medical risk data produced by IMM.

Development of insights from PRAs for non-PRA people

International Nuclear Information System (INIS)

Reilly, H.J.; Meale, B.M.

1992-01-01

A probabilistic risk assessment (PRA) of the Savannah River K-Reactor was completed in 1990. The PRA estimated the frequency of core damage accidents caused by operational occurrences during power operation of the reactor. The US Department of Energy (DOE) requested Idaho National Engineering Laboratory (INEL) to prepare guidance based on the PRA for use by DOE personnel at the Savannah River Site (SRS). The document had the purpose of informing the DOE system engineers and site representatives about how the information in the PRA might be used to help guide their activities. Opportunities existed to develop a document somewhat different than those developed previously by other programs. The opportunities existed because the audience is different: the principal audience for the document consists of DOE engineers who have continuing oversight responsibility for activities performed by the operating contractor at the K-Reactor, but who may not be knowledgeable about PRA

The role of PRA in the safety assessment of VVER Nuclear Power Plants in Ukraine

International Nuclear Information System (INIS)

Kot, C.

1999-01-01

Ukraine operates thirteen (13) Soviet-designed pressurized water reactors, VVERS. All Ukrainian plants are currently operating with annually renewable permits until they update their safety analysis reports (SARs), in accordance with new SAR content requirements issued in September 1995, by the Nuclear Regulatory Authority and the Government Nuclear Power Coordinating Committee of Ukraine. The requirements are in three major areas: design basis accident (DBA) analysis, probabilistic risk assessment (PRA), and beyond design-basis accident (BDBA) analysis. The last two requirements, on PRA and BDBA, are new, and the DBA requirements are an expanded version of the older SAR requirements. The US Department of Energy (USDOE), as part of its Soviet-Designed Reactor Safety activities, is providing assistance and technology transfer to Ukraine to support their nuclear power plants (NPPs) in developing a Western-type technical basis for the new SARs. USDOE sponsored In-Depth Safety Assessments (ISAs) are in progress at three pilot nuclear reactor units in Ukraine, South Ukraine Unit 1, Zaporizhzhya Unit 5, and Rivne Unit 1, and a follow-on study has been initiated at Khmenytskyy Unit 1. The ISA projects encompass most areas of plant safety evaluation, but the initial emphasis is on performing a detailed, plant-specific Level 1 Internal Events PRA. This allows the early definition of the plant risk profile, the identification of risk significant accident sequences and plant vulnerabilities and provides guidance for the remainder of the safety assessments

Practical Application of PRA as an Integrated Design Tool for Space Systems

Science.gov (United States)

Kalia, Prince; Shi, Ying; Pair, Robin; Quaney, Virginia; Uhlenbrock, John

2013-01-01

This paper presents the application of the first comprehensive Probabilistic Risk Assessment (PRA) during the design phase of a joint NASA/NOAA weather satellite program, Geostationary Operational Environmental Satellite Series R (GOES-R). GOES-R is the next generation weather satellite primarily to help understand the weather and help save human lives. PRA has been used at NASA for Human Space Flight for many years. PRA was initially adopted and implemented in the operational phase of manned space flight programs and more recently for the next generation human space systems. Since its first use at NASA, PRA has become recognized throughout the Agency as a method of assessing complex mission risks as part of an overall approach to assuring safety and mission success throughout project lifecycles. PRA is now included as a requirement during the design phase of both NASA next generation manned space vehicles as well as for high priority robotic missions. The influence of PRA on GOES-R design and operation concepts are discussed in detail. The GOES-R PRA is unique at NASA for its early implementation. It also represents a pioneering effort to integrate risks from both Spacecraft (SC) and Ground Segment (GS) to fully assess the probability of achieving mission objectives. PRA analysts were actively involved in system engineering and design engineering to ensure that a comprehensive set of technical risks were correctly identified and properly understood from a design and operations perspective. The analysis included an assessment of SC hardware and software, SC fault management system, GS hardware and software, common cause failures, human error, natural hazards, solar weather and infrastructure (such as network and telecommunications failures, fire). PRA findings directly resulted in design changes to reduce SC risk from micro-meteoroids. PRA results also led to design changes in several SC subsystems, e.g. propulsion, guidance, navigation and control (GNC

Use of PRA in the nuclear regulatory field in South Africa

International Nuclear Information System (INIS)

Hill, T.F.

1994-01-01

The nuclear regulatory authority in South Africa (since 1988 the Council for Nuclear Safety (CNS)), established in 1973 nuclear safety criteria against which to assess the level of safety of any facility using radioactive material. It is a regulatory requirement in South Africa to develop and maintain a living PRA for each facility and thereby to provide the necessary information to demonstrate compliance against these criteria. All safety submissions to the CNS must include at least a risk statement based on an accepted PRA study. The function of the CNS is to regulate all activities in South Africa involving the use of radioactive material and posing a significant risk to the public or plant personnel. This includes most aspects of the nuclear fuel cycle and the Koeberg NPS (two 2775 MW(th) PWRs). A PRA study including source terms for the two Koeberg units was presented by the contractor in 1979. This included the risk due to power and shutdown states and non reactor related accidents involving spent fuel storage, fuel handling and waste treatment related activities. At least 20 PRA studies have been performed for other nuclear facilities in the country. The CNS maintains an in-house PRA capability to perform independent assessments of licensee submission, to participate in developments of PRA methodology in the regulatory field, to perform pro-active safety work and to assist in regulatory decision making. Present ongoing work includes the development of a risk monitor, a risk management system, improvement in PRA codes, models, data collection and analysis, off-site risk assessment methodology and associated regulatory policy. (author). 1 fig

Use of probabilistic risk assessment (PRA) in expert systems to advise nuclear plant operators and managers

International Nuclear Information System (INIS)

Uhrig, R.E.

1988-01-01

The use of expert systems in nuclear power plants to provide advice to managers, supervisors and/or operators is a concept that is rapidly gaining acceptance. Generally, expert systems rely on the expertise of human experts or knowledge that has been codified in publications, books, or regulations to provide advice under a wide variety of conditions. In this work, a probabilistic risk assessment (PRA) of a nuclear power plant performed previously is used to assess the safety status of nuclear power plants and to make recommendations to the plant personnel. Nuclear power plants have many redundant systems and can continue to operate when one or more of these systems is disabled or removed from service for maintenance or testing. PRAs provide a means of evaluating the risk to the public associated with the operation of nuclear power plants with components or systems out of service. While the choice of the source term and methodology in a PRA may influence the absolute probability and consequences of a core melt, the ratio of the PRA calculations for two configurations of the same plant, carried out on a consistent basis, can readily identify the increase in risk associated with going from one configuration to the other

Constellation Probabilistic Risk Assessment (PRA): Design Consideration for the Crew Exploration Vehicle

Science.gov (United States)

Prassinos, Peter G.; Stamatelatos, Michael G.; Young, Jonathan; Smith, Curtis

2010-01-01

Managed by NASA's Office of Safety and Mission Assurance, a pilot probabilistic risk analysis (PRA) of the NASA Crew Exploration Vehicle (CEV) was performed in early 2006. The PRA methods used follow the general guidance provided in the NASA PRA Procedures Guide for NASA Managers and Practitioners'. Phased-mission based event trees and fault trees are used to model a lunar sortie mission of the CEV - involving the following phases: launch of a cargo vessel and a crew vessel; rendezvous of these two vessels in low Earth orbit; transit to th$: moon; lunar surface activities; ascension &om the lunar surface; and return to Earth. The analysis is based upon assumptions, preliminary system diagrams, and failure data that may involve large uncertainties or may lack formal validation. Furthermore, some of the data used were based upon expert judgment or extrapolated from similar componentssystemsT. his paper includes a discussion of the system-level models and provides an overview of the analysis results used to identify insights into CEV risk drivers, and trade and sensitivity studies. Lastly, the PRA model was used to determine changes in risk as the system configurations or key parameters are modified.

A PRA case study of extended long term decay heat removal for shutdown risk assessment

International Nuclear Information System (INIS)

Roglans, J.; Ragland, W.A.; Hill, D.J.

1992-01-01

A Probabilistic Risk Assessment (PRA) of the Experimental Breeder Reactor II (EBR-II), a Department of Energy (DOE) Category A research reactor, has recently been completed at Argonne National Laboratory (ANL). The results of this PRA have shown that the decay heat removal system for EBR-II is extremely robust and reliable. In addition, the methodology used demonstrates how the actions of other systems not normally used for actions of other systems not normally used for decay heat removal can be used to expand the mission time of the decay heat removal system and further increase its reliability. The methodology may also be extended to account for the impact of non-safety systems in enhancing the reliability of other dedicated safety systems

Interaction of CREDO [Centralized Reliability Data Organization] with the EBR-II [Experimental Breeder Reactor II] PRA [probabilistic risk assessment] development

International Nuclear Information System (INIS)

Smith, M.S.; Ragland, W.A.

1989-01-01

The National Academy of Sciences review of US Department of Energy (DOE) class 1 reactors recommended that the Experimental Breeder Reactor II (EBR-II), operated by Argonne National Laboratory (ANL), develop a level 1 probabilistic risk assessment (PRA) and make provisions for level 2 and level 3 PRAs based on the results of the level 1 PRA. The PRA analysis group at ANL will utilize the Centralized Reliability Data Organization (CREDO) at Oak Ridge National Laboratory to support the PRA data needs. CREDO contains many years of empirical liquid-metal reactor component data from EBR-II. CREDO is a mutual data- and cost-sharing system sponsored by DOE and the Power Reactor and Nuclear Fuels Development Corporation of Japan. CREDO is a component based data system; data are collected on components that are liquid-metal specific, associated with a liquid-metal environment, contained in systems that interface with liquid-metal environments, or are safety related for use in reliability/availability/maintainability (RAM) analyses of advanced reactors. The links between the EBR-II PRA development effort and the CREDO data collection at EBR-II extend beyond the sharing of data. The PRA provides a measure of the relative contribution to risk of the various components. This information can be used to prioritize future CREDO data collection activities at EBR-II and other sites

The Angra 1 fire PRA project

International Nuclear Information System (INIS)

Silva, Luiz E. Massiere de C.; Kassawara, Robert

2009-01-01

The Angra 1 Fire PRA (Probabilistic Risk Assessment) is under development by ELETRONUCLEAR jointly with EPRI (Electric Power Research Institute). The project was started January of 2007 and it is foreseen to be finished in the middle of the next year. The study is being conducted according to the newest methodology developed by EPRI and NRC/RES (U.S. Nuclear Regulatory Commission - Office of Regulatory Research) published in 2005 as Fire PRA Methodology for Nuclear Power Facilities (NUREG/CR-6850 or EPRI TR-1011989) [1]. Starting from the Internal Events Angra 1 PRA model Level 1 the project aims to be a comprehensive plant-specific fire analysis to identify the possible consequences of a fire in the plant vital areas which threaten the integrity of systems relevant to the safety, challenging the safety functions and representing a risk of accident that can lead to a core damage. The main tasks include the plant boundary and partitioning, the fire PRA component selection and the identification of the possible fire scenarios (ignition, propagation, detection, extinction and hazards) considering human failure events to establish the fire-induced risk model for quantification of the risk for nuclear core damage taking into account the plant design and its fire protection resources. This work presents a general discussion on the methodology applied to the completed steps of the project. (author)

'Living PRA' concept for plant risk: Reliability and availability tracking

International Nuclear Information System (INIS)

Sancaktar, S.; Sharp, D.R.

1985-01-01

The 'Living PRA' (Probabilistic Risk Assessment) is based on placing a PRA plant model on an interactive computer. This model consists of fault tree analyses for plant systems, event tree analyses for abnormal events and site specific consequence analysis for public and/or financial risks, for a nuclear power plant. A living PRA allows updates and sensitivity analyses by the plant owner throughout the lifetime of a plant. Recently, event and fault trees from two major PRAs were placed in a computerized format. The BYRON PRA study and the Living PRA and Economic Risk examples for Indian Point Unit-3 enabled analysts to gain experience and insight into the problems of plant operation. The above concept is well established for the Nuclear Power Plant evaluation. It has been also used for evaluation of processing facilities. In these studies, systems modeling was carried out by using the GRAFTER system for automated fault tree construction. Presently both the tools and the experience exists to set up useful and viable living PRA models for nuclear and chemical processing plants to enhance risk management by the plant owners through in-house use of micro computer based models

Chinshan living PRA model using NUPRA software package

International Nuclear Information System (INIS)

Cheng, S.-K.; Lin, T.-J.

2004-01-01

A living probabilistic risk assessment (PRA) model has been established for Chinshan Nuclear Power Station (BWR-4, MARK-I) using NUPRA software package. The core damage frequency due to internal events, seismic events and typhoons are evaluated in this model. The methodology and results considering the recent implementation of the 5th emergency diesel generator and automatic boron injection function are presented. The dominant sequences of this PRA model are discussed, and some possible applications of this living model are proposed. (author)

Level 2 PRA for a German BWR

International Nuclear Information System (INIS)

Sassen, F.; Rapp, W.; Tietsch, W.; Roess, P.

2007-01-01

A concept for a Level 2 Probabilistic Risk Assessment (L2 PRA) for a German Boiling Water Reactor (BWR) has been developed taking into account the role of L2 PRA within the German regulatory landscape. According to this concept, a plant specific evaluation of the severe accident phenomenology as well as analyses of the accident progression for the severe accident scenarios has been performed. Furthermore a plant specific MELCOR 1.8.6 model has been developed and special MELCOR source term calculations have been performed for the different release paths. This paper will present examples from the different areas described above. (author)

A comparative evaluation of five human reliability assessment techniques

International Nuclear Information System (INIS)

Kirwan, B.

1988-01-01

A field experiment was undertaken to evaluate the accuracy, usefulness, and resources requirements of five human reliability quantification techniques (Techniques for Human Error Rate Prediction (THERP); Paired Comparisons, Human Error Assessment and Reduction Technique (HEART), Success Liklihood Index Method (SLIM)-Multi Attribute Utility Decomposition (MAUD), and Absolute Probability Judgement). This was achieved by assessing technique predictions against a set of known human error probabilities, and by comparing their predictions on a set of five realistic Probabilisitc Risk Assessment (PRA) human error. On a combined measure of accuracy THERP and Absolute Probability Judgement performed best, whilst HEART showed indications of accuracy and was lower in resources usage than other techniques. HEART and THERP both appear to benefit from using trained assessors in order to obtain the best results. SLIM and Paired Comparisons require further research on achieving a robust calibration relationship between their scale values and absolute probabilities. (author)

Role of PRA in new NPP projects

International Nuclear Information System (INIS)

Julin, A.; Sandberg, J.; Virolainen, R.

2012-01-01

In Finland, a plant specific, Level 1 and 2 Probabilistic Risk Analysis (PRA) is required as a prerequisite for issuing the construction license and operating license. The use of PRA in various applications and the main insights are presented. These applications include e.g. PRA support to the design of SSCs (Systems, Structures and Components), definition of pre-service and in-service inspection programs, evaluation of the safety classification of SSCs, development of procedures, training and in definition of risk informed technical specifications, periodic testing and on-line preventive maintenance programs. In addition, PRA shall be used to assess the adequacy and coverage of the phase and system commissioning programs. Also the potential risks related to commissioning tests during nuclear test phase, shall be assessed with the help of PRA. In OL3 project, risk informed approach has been applied on a large scale for the first time in the design, construction and commissioning of a new NPP unit. Pre-nuclear commissioning tests have started at OL3 site and the plant is foreseen to begin commercial operation in 2013. Decisions have been made to launch new NPP projects. Teollisuuden Voima Oyj (TVO) is planning to build a new unit (OL4) at Olkiluoto site and a new utility, Fennovoima, is planning to build one unit at one of two alternative green field sites in Northern parts of Finland. Insights from PRAs of operating NPPs have been used in the evaluation of possible new sites to ensure that the site specific concerns and environmental conditions are adequately taken into account in the design of SSCs. Although the seismic activity at the Olkiluoto site is low, a comprehensive seismic risk analysis is being conducted. Its results support the review of the deterministic seismic design. For new sites, a probabilistic seismic hazard analysis has been carried out for the determination of the design earthquake. Experiences from OL3 licensing have been utilized in the

Issues and insights of PRA methodology in nuclear and space applications

International Nuclear Information System (INIS)

Hsu, F.

2005-01-01

This paper presents some important issues and technical insights on the scope, conceptual framework, and essential elements of nuclear power plant Probabilistic Risk Assessments (PRAs) and that of the PRAs in general applications of the aerospace industry, such as the Space Shuttle PRA being conducted by NASA. Discussions are focused on various lessons learned in nuclear power plant PRA applications and their potential applicability to the PRAs in the aerospace and launch vehicle systems. Based on insights gained from PRA projects for nuclear power plants and from the current Space Shuttle PRA effort, the paper explores the commonalities and the differences between the conduct of the different PRAs and the key issues and risk insights derived from extensive modeling practices in both industries of nuclear and space. (author)

Task analysis: How far are we from usable PRA input

International Nuclear Information System (INIS)

Gertman, D.I.; Blackman, H.S.; Hinton, M.F.

1984-01-01

This chapter reviews data collected at the Idaho National Engineering Laboratory for three DOE-owned reactors (the Advanced Test Reactor, the Power Burst Facility, and the Loss of Fluids Test Reactor) in order to identify usable Probabilistic Risk Assessment (PRA) input. Task analytic procedures involve the determination of manning and skill levels as a means of determining communication requirements, in assessing job performance aids, and in assessing the accuracy and completeness of emergency and maintenance procedures. The least understood aspect in PRA and plant reliability models is the human factor. A number of examples from the data base are discussed and offered as a means of providing more meaningful data than has been available to PRA analysts in the past. It is concluded that the plant hardware-procedures-personnel interfaces are essential to safe and efficient plant operations and that task analysis is a reasonably sound way of achieving a qualitative method for identifying those tasks most strongly associated with task difficulty, severity of consequence, and error probability

IRIS PRA preliminary results and future direction

International Nuclear Information System (INIS)

Finnicum, D.J.; Kling, C.L.; Carelli, M.D.

2004-01-01

Westinghouse is currently conducting the pre-application licensing of the International Reactor Innovative and Secure (IRIS) on behalf of the IRIS Consortium. One of the key aspects of the IRIS design is the concept of safety-by-design. The PRA (Probabilistic Risk Analysis) is being used as an integral part of the design process. As part of this effort, a PRA of the initial design was generated to address 2 key areas. First, the IRIS PRA supported the evaluation of IRIS design issues by providing a solid risk basis for design and analyses required for the pre-licensing evaluation of the IRIS design. The PRA provides the tool for quantifying the benefit of the safety-by-design approach. Second, the current PRA task is beginning the preparation of the more complete PRA analyses and documentation eventually required for Design Certification. One of the key risk-related goals for IRIS is to reduce the EPZ (Emergency Protection Zone) to within the exclusion area by demonstrating that the off-site doses are consistent with the US Protective Action Guidelines (PAGs) for initiation of emergency response so that the required protective actions would be limited to the exclusion area. The results of the preliminary PRA indicated a core damage frequency of 1.2 E-08 for internal initiators. This is a very good result but much work is needed to meet the ambitious goal of no emergency response. The next phase of the PRA analyses will involve a two-fold expansion of the PRA. First, as the design and analyses approach a greater level of detail, the assumptions used for the initial PRA will be reviewed and the models will be revised as needed to reflect the improved knowledge of the system design and performance. Furthermore, as the full plant design advances, the PRA will be expanded to incorporate risk associated with external challenges such as seismic and fire, and to address low power and shutdowns modes of operation. As with the initial work, the PRA will serve as a tool to

Development of margin assessment methodology of decay heat removal function against external hazards. (2) Tornado PRA methodology

International Nuclear Information System (INIS)

Nishino, Hiroyuki; Kurisaka, Kenichi; Yamano, Hidemasa

2014-01-01

Probabilistic Risk Assessment (PRA) for external events has been recognized as an important safety assessment method after the TEPCO's Fukushima Daiichi nuclear power station accident. The PRA should be performed not only for earthquake and tsunami which are especially key events in Japan, but also the PRA methodology should be developed for the other external hazards (e.g. tornado). In this study, the methodology was developed for Sodium-cooled Fast Reactors paying attention to that the ambient air is their final heat sink for removing decay heat under accident conditions. First, tornado hazard curve was estimated by using data recorded in Japan. Second, important structures and components for decay heat removal were identified and an event tree resulting in core damage was developed in terms of wind load and missiles (i.e. steel pipes, boards and cars) caused by a tornado. Main damage cause for important structures and components is the missiles and the tornado missiles that can reach those components and structures placed on high elevations were identified, and the failure probabilities of the components and structures against the tornado missiles were calculated as a product of two probabilities: i.e., a probability for the missiles to enter the intake or outtake in the decay heat removal system, and a probability of failure caused by the missile impacts. Finally, the event tree was quantified. As a result, the core damage frequency was enough lower than 10 -10 /ry. (author)

PRA research and the development of risk-informed regulation at the U.S. nuclear regulatory commission

International Nuclear Information System (INIS)

Siu, Nathan; Collins, Dorothy

2008-01-01

Over the years, Probabilistic Risk Assessment (PRA) research activities conducted at the U.S. Nuclear Regulatory Commission (NRC) have played an essential role in support of the agency's move towards risk-informed regulation. These research activities have provided the technical basis for NRC's regulatory activities in key areas; provided PRA methods, tools, and data enabling the agency to meet future challenges; supported the implementation of NRC's 1995 PRA Policy Statement by assessing key sources of risk; and supported the development of necessary technical and human resources supporting NRC's risk-informed activities. PRA research aimed at improving the NRC's understanding of risk can positively affect the agency's regulatory activities, as evidenced by three case studies involving research on fire PRA, Human Reliability Analysis (HRA), and Pressurized Thermal Shock (PTS) PRA. These case studies also show that such research can take a considerable amount of time, and that the incorporation of research results into regulatory practice can take even longer. The need for sustained effort and appropriate lead time is an important consideration in the development of a PRA research program aimed at helping the agency address key sources of risk for current and potential future facilities

Development of a methodology for conducting an integrated HRA/PRA --

Energy Technology Data Exchange (ETDEWEB)

Luckas, W.J.; Barriere, M.T.; Brown, W.S. (Brookhaven National Lab., Upton, NY (United States)); Wreathall, J. (Wreathall (John) and Co., Dublin, OH (United States)); Cooper, S.E. (Science Applications International Corp., McLean, VA (United States))

1993-01-01

During Low Power and Shutdown (LP S) conditions in a nuclear power plant (i.e., when the reactor is subcritical or at less than 10--15% power), human interactions with the plant's systems will be more frequent and more direct. Control is typically not mediated by automation, and there are fewer protective systems available. Therefore, an assessment of LP S related risk should include a greater emphasis on human reliability than such an assessment made for power operation conditions. In order to properly account for the increase in human interaction and thus be able to perform a probabilistic risk assessment (PRA) applicable to operations during LP S, it is important that a comprehensive human reliability assessment (HRA) methodology be developed and integrated into the LP S PRA. The tasks comprising the comprehensive HRA methodology development are as follows: (1) identification of the human reliability related influences and associated human actions during LP S, (2) identification of potentially important LP S related human actions and appropriate HRA framework and quantification methods, and (3) incorporation and coordination of methodology development with other integrated PRA/HRA efforts. This paper describes the first task, i.e., the assessment of human reliability influences and any associated human actions during LP S conditions for a pressurized water reactor (PWR).

Overview of the probabilistic risk assessment approach

International Nuclear Information System (INIS)

Reed, J.W.

1985-01-01

The techniques of probabilistic risk assessment (PRA) are applicable to Department of Energy facilities. The background and techniques of PRA are given with special attention to seismic, wind and flooding external events. A specific application to seismic events is provided to demonstrate the method. However, the PRA framework is applicable also to wind and external flooding. 3 references, 8 figures, 1 table

Emergence of Participatory Rural Appraisal (PRA Technique as a Strategy towards Sustainable Development: A Sri Lankan Experience

Directory of Open Access Journals (Sweden)

D. N. Koralagama

2007-10-01

Full Text Available In this millennium all the development activities are mostly focused on sustainable development, i.e. the development which fulfils the requirements of the present without disturbing the utilization of future generation. Basically, the sustainable development deals with environmental, social, and economical initiations. In relation to these three objectives, community participation plays a key role as an effective strategy for sustainable development. Among the numerous types of participation, Participatory Rural Appraisal (PRA technique is the most relevant effective method to receive the participation. Because, it has been strengthen by bottom up approach, well defined objectives, practicable solutions, and remedies. Hence, the out come of such an event is most productive rather than a top bottom approach techniques. In fact, a PRA was practiced to develop a strategic plan for tsunami affected village – Bambaranda east, in southern province of Sri Lanka. PRA sessions were carried out during February, 2007 by the Department of Agric. Economics of Ruhuna University, Sri Lanka in collaboration with Japanese Green Resource Agency, Japan.Participatory mapping, venn diagram, matrix ranking, preference ranking, and pair - wise ranking were demonstrated to gather information from the community. The tsunami affected area, including the paddy fields, four irrigation canals were shown by the group with the help of the participatory map. Preference ranking was resulted the reconstruction of irrigation canals as the most important rehabilitation activity to recover the livelihood of villagers. Intrusion of sea water into the paddy fields was the main limitation revealed by the pair - wise ranking. The second limitation marked as unavailability of enough fertilizer and the dilapidated irrigation canals was the third that has to be solved. Matrix ranking was employed to identify the most facilitated sectors by the government and other institutes in order

Certification plan for safety and PRA codes

International Nuclear Information System (INIS)

Toffer, H.; Crowe, R.D.; Ades, M.J.

1990-05-01

A certification plan for computer codes used in Safety Analyses and Probabilistic Risk Assessment (PRA) for the operation of the Savannah River Site (SRS) reactors has been prepared. An action matrix, checklists, and a time schedule have been included in the plan. These items identify what is required to achieve certification of the codes. A list of Safety Analysis and Probabilistic Risk Assessment (SA ampersand PRA) computer codes covered by the certification plan has been assembled. A description of each of the codes was provided in Reference 4. The action matrix for the configuration control plan identifies code specific requirements that need to be met to achieve the certification plan's objectives. The checklist covers the specific procedures that are required to support the configuration control effort and supplement the software life cycle procedures based on QAP 20-1 (Reference 7). A qualification checklist for users establishes the minimum prerequisites and training for achieving levels of proficiency in using configuration controlled codes for critical parameter calculations

PRA and the implementation of quantitative safety goals

International Nuclear Information System (INIS)

Okrent, D.

1983-01-01

With the adoption by the U.S. Nuclear Regulatory Commission (NRC) in January, 1983, of a Policy Statement on Safety Goals for the Operation of Nuclear Power Plants, probabilitstic risk assessment (PRA) has taken on increased importance in nuclear reactor safety. Although the Reactor Safety Study, WASH-1400, was a major pioneering effort that revolutionized thinking about reactor safety, PRA was used only on occasion by the NRC regulatory staff prior to the accident at Three Mile Island. Since then, PRA has been used more and more as an important factor in decision making, usually for specific issues. The nuclear industry has also employed PRA, sometimes to make its case on specific issues, sometimes to present a position on overall risk. The advent of the Zion and Indian Point PRAs, with their treatment of risks from fire, wind, and earthquakes, and their examination of the course of core melt accidents, has added a new dimension to the overall picture. Although the NRC has stated that during the next two year evolution period, its quantitative design objectives and PRA are not to enter directly into the licensing process, many important issues will be influenced significantly by the results of risk and reliability studies. In fact, PRA may be coming into a position of great importance before the methodology, data, and process are sufficiently mature for the task. Large gaps still exist in our understanding of phenomena and in input information; and much of the final result depends on subjective input; large differences of opinion can and should be expected to persist. Accepted standards for quality assurance, and adequacy and depth of independent, peer review remain to be formulated and achieved. This paper will summarize the recently adopted NRC safety policy and the two-year evaluation plan, and will provide, by example, some words of caution concerning a few of the difficulties which may arise. (orig.)

Development of a methodology for conducting an integrated HRA/PRA --. Task 1, An assessment of human reliability influences during LP&S conditions PWRs

Energy Technology Data Exchange (ETDEWEB)

Luckas, W.J.; Barriere, M.T.; Brown, W.S. [Brookhaven National Lab., Upton, NY (United States); Wreathall, J. [Wreathall (John) and Co., Dublin, OH (United States); Cooper, S.E. [Science Applications International Corp., McLean, VA (United States)

1993-06-01

During Low Power and Shutdown (LP&S) conditions in a nuclear power plant (i.e., when the reactor is subcritical or at less than 10--15% power), human interactions with the plant`s systems will be more frequent and more direct. Control is typically not mediated by automation, and there are fewer protective systems available. Therefore, an assessment of LP&S related risk should include a greater emphasis on human reliability than such an assessment made for power operation conditions. In order to properly account for the increase in human interaction and thus be able to perform a probabilistic risk assessment (PRA) applicable to operations during LP&S, it is important that a comprehensive human reliability assessment (HRA) methodology be developed and integrated into the LP&S PRA. The tasks comprising the comprehensive HRA methodology development are as follows: (1) identification of the human reliability related influences and associated human actions during LP&S, (2) identification of potentially important LP&S related human actions and appropriate HRA framework and quantification methods, and (3) incorporation and coordination of methodology development with other integrated PRA/HRA efforts. This paper describes the first task, i.e., the assessment of human reliability influences and any associated human actions during LP&S conditions for a pressurized water reactor (PWR).

PRA -- Now that operators have it, what do they do with it?

International Nuclear Information System (INIS)

Rasmussen, M.A.; Kolo, R.J.

1996-01-01

Many utilities have had Probabilistic Risk Assessment (PRA) projects underway for several years in order to satisfy the NRC Generic Letter 88-20 requirement for an Individual Plant Examination, or IPE. Typically the studies have reached the conclusion that there are significant differences in the contribution of different plant components to preventing core damage should a major plant transient occur. How nuclear plant operators can use this knowledge to DECREASE the overall risk of performing the routine tasks of testing and maintenance is not an easy task. 10CFR50.65; ''The Maintenance Rule,'' requires that any plant maintenance performed with the unit on line be evaluated for risk. Byron Station will satisfy the 10CFR50.65 requirement by using PRA methodology to evaluate testing and maintenance activities performed with the unit at power. The challenge is to effectively use the results of PRA studies to aid in plant operations without having to make on shift plant operations personnel experts in PRA. At Byron, PRA is used to help build the weekly work schedules. Operations personnel tasked with reviewing the work schedule are the departmental experts on the use of the PRA results. The on shift SRO's role in implementing the program is to accurately execute and monitor the work week schedule as written, and to react to unforeseen equipment failures with an appropriate level of response. The response to such emergent work items is also predefined. Handling emergent work in a prescribed manner minimizes the overall risk to the unit and also eliminates the need to have PRA expertise available to make emergent work risk evaluations. Thus the on shift operators' required knowledge of PRA methods and intricacies is minimized. PRA is just another of the many tools used by the shift operator to run the plant in a safe, conservative manner

Development of a methodology for conducting an integrated HRA/PRA --

International Nuclear Information System (INIS)

Luckas, W.J.; Barriere, M.T.; Brown, W.S.; Wreathall, J.; Cooper, S.E.

1993-01-01

During Low Power and Shutdown (LP ampersand S) conditions in a nuclear power plant (i.e., when the reactor is subcritical or at less than 10--15% power), human interactions with the plant's systems will be more frequent and more direct. Control is typically not mediated by automation, and there are fewer protective systems available. Therefore, an assessment of LP ampersand S related risk should include a greater emphasis on human reliability than such an assessment made for power operation conditions. In order to properly account for the increase in human interaction and thus be able to perform a probabilistic risk assessment (PRA) applicable to operations during LP ampersand S, it is important that a comprehensive human reliability assessment (HRA) methodology be developed and integrated into the LP ampersand S PRA. The tasks comprising the comprehensive HRA methodology development are as follows: (1) identification of the human reliability related influences and associated human actions during LP ampersand S, (2) identification of potentially important LP ampersand S related human actions and appropriate HRA framework and quantification methods, and (3) incorporation and coordination of methodology development with other integrated PRA/HRA efforts. This paper describes the first task, i.e., the assessment of human reliability influences and any associated human actions during LP ampersand S conditions for a pressurized water reactor (PWR)

Use of plant-specific PRA in an EOP scope audit

International Nuclear Information System (INIS)

O'Brien, J.J.

1991-01-01

Traditionally, decisions on which accident scenarios to proceduralize as emergency operating procedures (EOPs) have been based on existing design basis analyses, engineering judgment, and probabilistic risk assessments (PRAs) on generic plants. This approach has important strengths and limits. The major limitation of generic PRAs is their inability to account for plant-specific features. Use of plant-specific PRA to determine the impact of proceduralizing, or not proceduralizing, responses to scenarios considers plant-specific features. This helps to eliminate unnecessary EOPs, thus allowing resources to be concentrated on scenarios that are more important for a particular plant. In preparation for a US Nuclear Regulatory Commission audit, a plant-specific PRA was used to assess and quantify the plant's previous decision not to implement six reference emergency response guidelines (ERGs) as procedures. The original justification for nonimplementation of the ERGs was based on engineering judgment. The PRA provided a quantitative justification for implementation/nonimplementation of each guidelines. This analysis accounted for plant-specific design features not common to all reference plants

Selecting the seismic HRA approach for Savannah River Plant PRA revision 1

International Nuclear Information System (INIS)

Papouchado, K.; Salaymeh, J.

1993-10-01

The Westinghouse Savannah River Company (WSRC) has prepared a level I probabilistic risk assessment (PRA), Rev. 0 of reactor operations for externally-initiated events including seismic events. The SRS PRA, Rev. 0 Seismic HRA received a critical review that expressed skepticism with the approach used for human reliability analysis because it had not been previously used and accepted in other published PRAs. This report provides a review of published probabilistic risk assessments (PRAs), the associated methodology guidance documents, and the psychological literature to identify parameters important to seismic human reliability analysis (HRA). It also describes a recommended approach for use in the Savannah River Site (SRS) PRA. The SRS seismic event PRA performs HRA to account for the contribution of human errors in the accident sequences. The HRA of human actions during and after a seismic event is an area subject to many uncertainties and involves significant analyst judgment. The approach recommended by this report is based on seismic HRA methods and associated issues and concerns identified from the review of these referenced documents that represent the current state-of-the- art knowledge and acceptance in the seismic HRA field

PRA: a powerful engineering decision tool

International Nuclear Information System (INIS)

Carvalho, H.G. de.

1988-03-01

The probabilistic risk analysis (PRA) is studied and its historical development is briefly presented. Human factors, sofware and guides, improvement of utility management of nuclear power operations are discussed. The development of a standardized LWR design, optimized for safety, reliability and economy is studied. The impact of risk assessments in public acceptance of nuclear power is discussed. (M.A.C.) [pt

PRA and Conceptual Design

Science.gov (United States)

DeMott, Diana; Fuqua, Bryan; Wilson, Paul

2013-01-01

Once a project obtains approval, decision makers have to consider a variety of alternative paths for completing the project and meeting the project objectives. How decisions are made involves a variety of elements including: cost, experience, current technology, ideologies, politics, future needs and desires, capabilities, manpower, timing, available information, and for many ventures management needs to assess the elements of risk versus reward. The use of high level Probabilistic Risk Assessment (PRA) Models during conceptual design phases provides management with additional information during the decision making process regarding the risk potential for proposed operations and design prototypes. The methodology can be used as a tool to: 1) allow trade studies to compare alternatives based on risk, 2) determine which elements (equipment, process or operational parameters) drives the risk, and 3) provide information to mitigate or eliminate risks early in the conceptual design to lower costs. Creating system models using conceptual design proposals and generic key systems based on what is known today can provide an understanding of the magnitudes of proposed systems and operational risks and facilitates trade study comparisons early in the decision making process. Identifying the "best" way to achieve the desired results is difficult, and generally occurs based on limited information. PRA provides a tool for decision makers to explore how some decisions will affect risk before the project is committed to that path, which can ultimately save time and money.

Value impact analysis utilizing PRA techniques combined with a hybrid plant model

International Nuclear Information System (INIS)

Edson, J.L.; Stillwell, D.W.

1989-01-01

A value impact analysis (VIA) has been performed by the INEL to support a NRC Regulatory Analysis for resolution of Generic Issue (GI) 29, Bolting Degradation or Failure in Nuclear Power Plants. A VIA for replacing the reactor coolant pressure boundary (RCPB) bolts of BWRs and PWRs was previously prepared by Pacific Northwest Laboratories in 1985 under instructions limiting the VIA to the potential for failure of primary pressure boundary bolting. Subsequently the INEL was requested to perform a VIA that included non primary systems and component support bolts to be compatible with the resolution of the broader issue. Because the initial list of systems and bolting applications that could be included in the VIA was very large, including them all in the VIA would likely result in analyzing some that have little if any effect on public risk. This paper discusses how PRA techniques combined with a hybrid plant model were used to determine which bolts have the potential to be significant contributors to public risk if they were to fail, and therefore were included in the VIA

MAAP4.0.7 analysis and justification for PRA level 1 mission success criteria

International Nuclear Information System (INIS)

Butler, J.S.; Kapitz, D.; Martin, R.P.; Seifaee, F.; Sundaram, R.K.

2008-01-01

The U.S. EPR is a 4590 MWth evolutionary pressurized water reactor that incorporates proven technology with innovative system architecture to provide an unprecedented level of safety. One of the measures of safety is provided by Probability Risk Assessment (PRA). PRA Level 1 concerns the evaluation of core damage frequency based on various initiating events and the success or failure of various plant event mitigation features. Determination of this measure requires mission success criteria, which are used to build the logic that makes up the fault trees and event trees of the Level 1 PRA. Developing mission success criteria for the wide variety of accident sequences modeled in the PRA Level 1 model requires a large number of thermal hydraulic calculations. The MAAP4 code, developed by Fauske and Associates, Inc. and distributed by EPRI, was chosen to perform these calculations because of its fast computation times relative to more sophisticated thermal-hydraulics codes This is a unique application of MAAP4, which was developed specifically for severe accident and PRA Level 2 analysis. As such, a study was performed to assess MAAP4 's thermal-hydraulic response capabilities against AREVA 's S-RELAP5 best-estimate integral systems thermal-hydraulic analysis code. (authors)

Observations on PRA and its applications

International Nuclear Information System (INIS)

Yeh, Y.-C.; Shieh, S.-L.

2004-01-01

An overview on the experience of PRA and its prospective application in Taiwan's three nuclear power plants is presented. Through the PRA, plant design improvements are performed and several engineering findings are illuminated. The sensitivity study including the internal, seismic, and typhoon events are conducted to justify items that can significantly reduce core meltdown risk. Its resulted plant betterment plans are thus highlighted accordingly. For PRA application, a risk-based inspection program for allocating inspection human resources has been resulted following the importance ranking of each component. The developing risk-based regulation to rationalize technical specification and maintenance program will also be entailed. To enhance the accuracy of the PRA model and its reproducibility, several issues are considered to have high priority for improvement such as external event data and analyses, uncertainty, common mode failure, human reliability, and the relative component importance. Highlight of their significance along with some typical sensitivity analyses are discussed for further investigation. (author)

Use of PRA techniques to optimize the design of the IRIS nuclear power plant

International Nuclear Information System (INIS)

Muhlheim, M.D.; Cletcher, J.W. II

2003-01-01

True design optimization of a plants inherent safety and performance characteristics results when a probabilistic risk assessment (PRA) is integrated with the plant-level design process. This is the approach being used throughout the design of the International Reactor Innovative and Secure (IRIS) nuclear power plant to maximize safety. A risk-based design optimization tool employing a 'one-button' architecture is being developed by the Oak Ridge National Laboratory to evaluate design changes; new modeling approaches, methods, or theories modeling uncertainties and completeness; physical assumptions; and data changes on component, cabinet, train, and system bases. Unlike current PRAs, the one-button architecture allows components, modules, and data to be interchanged at will with the probabilistic effect immediately apparent. Because all of the current and previous design, and data sets are available via the one-button architecture, the safety ramifications of design options are evaluated, feedback on design alternatives is immediate, and true optimization and understanding can be achieved. Thus, for the first time, PRA analysts and designers can easily determine the probabilistic implications of different design configurations and operating conditions in various combinations for the entire range of initiating events. The power of the one-button architecture becomes evident by the number of design alternatives that can be evaluated C11 component choices yielded 160 design alternatives. Surprisingly, the lessons learned can be counter-intuitive and significant. For example, one of the alternative designs for IRIS evaluated via this architecture revealed that because of common-cause failure probabilities, using the most reliable components actually decreased systems' reliability. (author)

PRA quality and use

International Nuclear Information System (INIS)

Okrent, D.; Apostolakis, G.; Whitley, R.; Garrick, B.J.

1982-10-01

This report deals with several inter-related aspects of probabilistic risk assessment. Some prior opinion regarding quality assurance, methodology and questions of peer review are reviewed, followed by comments by the authors on these and related subjects. Problems arising in decision-making by different groups concerning the meaning and validity of a PRA are examined, and the role of performance criteria in helping to achieve consensus is treated. Finally, a general approach to the development of performance criteria for systems and functions by the retrospective comparison of existing PRAs is proposed and examined in a preliminary fashion

Use of PRA in Shuttle Decision Making Process

Science.gov (United States)

Boyer, Roger L.; Hamlin, Teri L.

2010-01-01

How do you use PRA to support an operating program? This presentation will explore how the Shuttle Program Management has used the Shuttle PRA in its decision making process. It will reveal how the PRA has evolved from a tool used to evaluate Shuttle upgrades like Electric Auxiliary Power Unit (EAPU) to a tool that supports Flight Readiness Reviews (FRR) and real-time flight decisions. Specific examples of Shuttle Program decisions that have used the Shuttle PRA as input will be provided including how it was used in the Hubble Space Telescope (HST) manifest decision. It will discuss the importance of providing management with a clear presentation of the analysis, applicable assumptions and limitations, along with estimates of the uncertainty. This presentation will show how the use of PRA by the Shuttle Program has evolved overtime and how it has been used in the decision making process providing specific examples.

System 80+TM PRA insights on severe accident prevention and mitigation

International Nuclear Information System (INIS)

Finnicum, D.J.; Jacob, M.C.; Schneider, R.E.; Weston, R.A.

2004-01-01

The System 80 + design is ABB-CE's standardized evolutionary Advanced Light Water Reactor (ALWR) design. It incorporates design enhancements based on Probabilistic Risk Assessment (PRA) insights, guidance from the ALWR Utility Requirements Document (URD), and US NRC's Severe Accident Policy. Major severe accident prevention and mitigation design features of the System 80 + design are described. The results of the System 80 + PRA are presented and the insights gained from the PRA sensitivity analyses are discussed. ABB-CE considered defense-in-depth for accident prevention and mitigation early in the design process and used robust design features to ensure that the System 80 + design achieved a low core damage frequency, low containment conditional failure probability, and excellent deterministic containment performance under severe accident conditions and to ensure that the risk was properly allocated among design features and between prevention and mitigation. (author)

Recovery actions in PRA [probabilistic risk assessment] for the Risk Methods Integration and Evaluation Program (RMIEP): Volume 1, Development of the data-based method

International Nuclear Information System (INIS)

Weston, L.M.; Whitehead, D.W.; Graves, N.L.

1987-06-01

In a probabilistic risk assessment (PRA) for a nuclear power plant, the analyst identifies a set of potential core damage events consisting of equipment failures and human errors and their estimated probabilities of occurrence. If operator recovery from an event within some specified time is considered, then the probability of this recovery can be included in the PRA. This report provides PRA analysts with an improved methodology for including recovery actions in a PRA. A recovery action can be divided into two distinct phases: a Diagnosis Phase (realizing that there is a problem with a critical parameter and deciding upon the correct course of action) and an Action Phase (physically accomplishing the required action). In this methodology, simulator data are used to estimate recovery probabilities for the diagnosis phase. Different time-reliability curves showing the probability of failure of diagnosis as a function of time from the compelling cue for the event are presented. These curves are based on simulator exercises, and the actions are grouped based upon their operational similarities. This is an improvement over existing diagnosis models that rely greatly upon subjective judgment to obtain such estimates. The action phase is modeled using estimates from available sources. The methodology also includes a recommendation on where and when to apply the recovery action in the PRA process

Preliminary ATWS analysis for the IRIS PRA

International Nuclear Information System (INIS)

Maddalena Barra; Marco S Ghisu; David J Finnicum; Luca Oriani

2005-01-01

Full text of publication follows: The pressurized light water cooled, medium power (1000 MWt) IRIS (International Reactor Innovative and Secure) has been under development for four years by an international consortium of over 21 organizations from ten countries. The plant conceptual design was completed in 2001 and the preliminary design is nearing completion. The pre-application licensing process with NRC started in October, 2002. IRIS has been primarily focused on establishing a design with innovative safety characteristics. The first line of defense in IRIS is to eliminate event initiators that could potentially lead to core damage. In IRIS, this concept is implemented through the 'safety by design' approach, which allows to minimize the number and complexity of the safety systems and required operator actions. The end result is a design with significantly reduced complexity and improved operability, and extensive plant simplifications to enhance construction. To support the optimization of the plant design and confirm the effectiveness of the safety by design approach in mitigating or eliminating events and thus providing a significant reduction in the probability of severe accidents, the PRA is being used as an integral part of the design process. A preliminary but extensive Level 1 PRA model has been developed to support the pre-application licensing of the IRIS design. As a result of the Preliminary IRIS PRA, an optimization of the design from a reliability point of view was completed, and an extremely low (about 1.2 E -8 ) core damage frequency (CDF) was assessed to confirm the impact of the safety by design approach. This first assessment is a result of a PRA model including internal initiating events. During this assessment, several assumptions were necessary to complete the CDF evaluation. In particular Anticipated Transients Without Scram (ATWS) were not included in this initial assessment, because their contribution to core damage frequency was assumed

Use of probabilistic risk assessment (PRA) in expert systems to advise nuclear plant operators and managers

International Nuclear Information System (INIS)

Uhrig, R.E.

1988-01-01

The use of expert systems in nuclear power plants to provide advice to managers, supervisors and/or operators is a concept that is rapidly gaining acceptance. Generally, expert systems rely on the expertise of human experts or knowledge that has been modified in publications, books, or regulations to provide advice under a wide variety of conditions. In this work, a probabilistic risk assessment (PRA) 3 of a nuclear power plant performed previously is used to assess the safety status of nuclear power plants and to make recommendations to the plant personnel. 5 refs., 1 fig., 2 tabs

How Can You Support RIDM/CRM/RM Through the Use of PRA

Science.gov (United States)

DoVemto. Tpmu

2011-01-01

Probabilistic Risk Assessment (PRA) is one of key Risk Informed Decision Making (RIDM) tools. It is a scenario-based methodology aimed at identifying and assessing Safety and Technical Performance risks in complex technological systems.

Seismic PRA of a BWR plant

International Nuclear Information System (INIS)

Nishio, Masahide; Fujimoto, Haruo

2014-01-01

Since the occurrence of nuclear power plant accidents in the Fukushima Daichi nuclear power station, the regulatory framework on severe accident (SA) has been discussed in Japan. The basic concept is to typify and identify the accident sequences leading to core/primary containment vessel (PCV) damage and to implement SA measures covering internal and external events extensively. As Japan is an earthquake-prone country and earthquakes and tsunami are important natural external events for nuclear safety of nuclear power plants, JNES performed the seismic probabilistic risk assessment (PRA) on a typical nuclear power plant and evaluated the dominant accident sequences leading to core/PCV damage to discuss dominant scenarios of severe accident (SA). The analytical models and the results of level-1 seismic PRA on a 1,100 MWe BWR-5 plant are shown here. Seismic PRA was performed for a typical BWR5 plant. Initiating events with large contribution to core damage frequency are the loss of all AC powers (station blackout) and the large LOCA. The top of dominant accident sequences is the simultaneous occurrence of station blackout and large LOCA. Important components to core damage frequency are electric power supply equipment. It needs to keep in mind that the results are influenced on site geologic characteristic to a greater or lesser. In the process of analysis, issues such as conservative assumptions related to damages of building or structure and success criteria for excessive LOCA are left to be resolved. These issues will be further studied including thermal hydric analysis in the future. (authors)

Using level-I PRA for enhanced safety of the advanced neutron source reactor

International Nuclear Information System (INIS)

Ramsey, C.T.; Linn, M.A.

1995-01-01

The phase-1, level-I probabilistic risk assessment (PRA) of the Advanced Neutron Source (ANS) reactor has been completed as part of the conceptual design phase of this proposed research facility. Since project inception, PRA and reliability concepts have been an integral part of the design evolutions contributing to many of the safety features in the current design. The level-I PRA has been used to evaluate the internal events core damage frequency against project goals and to identify systems important to safety and availability, and it will continue to guide and provide support to accident analysis, both severe and nonsevere. The results also reflect the risk value of defense-in-depth safety features in reducing the likelihood of core damage

Calculation of Fire Severity Factors and Fire Non-Suppression Probabilities For A DOE Facility Fire PRA

International Nuclear Information System (INIS)

Elicson, Tom; Harwood, Bentley; Lucek, Heather; Bouchard, Jim

2011-01-01

Over a 12 month period, a fire PRA was developed for a DOE facility using the NUREG/CR-6850 EPRI/NRC fire PRA methodology. The fire PRA modeling included calculation of fire severity factors (SFs) and fire non-suppression probabilities (PNS) for each safe shutdown (SSD) component considered in the fire PRA model. The SFs were developed by performing detailed fire modeling through a combination of CFAST fire zone model calculations and Latin Hypercube Sampling (LHS). Component damage times and automatic fire suppression system actuation times calculated in the CFAST LHS analyses were then input to a time-dependent model of fire non-suppression probability. The fire non-suppression probability model is based on the modeling approach outlined in NUREG/CR-6850 and is supplemented with plant specific data. This paper presents the methodology used in the DOE facility fire PRA for modeling fire-induced SSD component failures and includes discussions of modeling techniques for: Development of time-dependent fire heat release rate profiles (required as input to CFAST), Calculation of fire severity factors based on CFAST detailed fire modeling, and Calculation of fire non-suppression probabilities.

Fire PRA requantification studies. Final report

International Nuclear Information System (INIS)

Parkinson, W.

1993-03-01

This report describes the requantification of two existing fire probabilistic risk assessments (PRAs) using a fire PRA method and data that are being developed by the Electric Power Research Institute (EPRI). The two existing studies are the Seabrook Station Probabilistic Safety Assessment that was made in 1983 and the 1989 NUREG-1150 analysis of the Peach Bottom Plant. Except for the fire methods and data, the original assumptions were used. The results from the requantification show that there were excessive conservatisms in the original studies. The principal reason for a hundredfold reduction in the Peach Bottom core- damage frequency is the determination that no electrical cabinet fire in a switchgear room would damage both offsite power feeds. Past studies often overestimated the heat release from electrical cabinet fires. EPRI's electrical cabinet heat release rates are based on tests that were conducted for Sandia's fire research program. The rates are supported by the experience in the EPRI Fire Events Database for U.S. nuclear plants. Test data and fire event experience also removed excessive conservatisms in the Peach Bottom control and cable spreading rooms, and the Seabrook primary component cooling pump, turbine building relay and cable spreading rooms. The EPRI fire PRA method and data will show that there are excessive conservatisms in studies that were made for many plants and can benefit them accordingly

Overview of seismic probabilistic risk assessment for structural analysis in nuclear facilities

International Nuclear Information System (INIS)

Reed, J.W.

1989-01-01

Probabilistic Risk Assessment (PRA) for seismic events is currently being performed for nuclear and DOE facilities. The background on seismic PRA is presented along with a basic description of the method. The seismic PRA technique is applicable to other critical facilities besides nuclear plants. The different approaches for obtained structure fragility curves are discussed and their applications to structures and equipment, in general, are addressed. It is concluded that seismic PRA is a useful technique for conducting probability analysis for a wide range of classes of structures and equipment

Individual plant examination and future PRA applications

International Nuclear Information System (INIS)

Monty, B.S.; Sursock, J.P.; Thierry, R.J.

1992-01-01

PRA is being used in many areas of plant operation as has been demonstrated in previous studies. With the U.S. NRC's emphasis on the use of risk to identify plant vulnerabilities and the development of plant specific PRA models for all plants, it is expected that the use of PRA will be expanded. Key areas where this is expected to occur include the development of risk-based Technical Specifications, risk management, and risk-centered maintenance programs. This paper focuses on the Individual Plant Examination requirement and the possible uses of risk-based methods in controlling plant operation to enhance plant safety and availability, and how the IPE requirement will potentially further this area of development. (orig./DG)

A perspective of PC-based probabilistic risk assessment

International Nuclear Information System (INIS)

Sattison, M.B.; Rasmuson, D.M.; Robinson, R.C.; Russell, K.D.; Van Siclen, V.S.

1987-01-01

Probabilistic risk assessment (PRA) information has been under-utilized in the past due to the large effort required to input the PRA data and the large expense of the computers needed to run PRA codes. The microcomputer-based Integrated Reliability and Risk Analysis System (IRRAS) and the System Analysis and Risk Assessment (SARA) System, under development at the Idaho National Engineering Laboratory, have greatly enhanced the ability of managers to use PRA techniques in their decision-making. IRRAS is a tool that allows an analyst to create, modify, update, and reanalyze a plant PRA to keep the risk assessment current with the plant's configuration and operation. The SARA system is used to perform sensitivity studies on the results of a PRA. This type of analysis can be used to evaluate proposed changes to a plant or its operation. The success of these two software projects demonstrate that risk information can be made readily available to those that need it. This is the first step in the development of a true risk management capability

Development of risk assessment methodology against natural external hazards for sodium-cooled fast reactors: project overview and strong Wind PRA methodology - 15031

International Nuclear Information System (INIS)

Yamano, H.; Nishino, H.; Kurisaka, K.; Okano, Y.; Sakai, T.; Yamamoto, T.; Ishizuka, Y.; Geshi, N.; Furukawa, R.; Nanayama, F.; Takata, T.; Azuma, E.

2015-01-01

This paper describes mainly strong wind probabilistic risk assessment (PRA) methodology development in addition to the project overview. In this project, to date, the PRA methodologies against snow, tornado and strong wind were developed as well as the hazard evaluation methodologies. For the volcanic eruption hazard, ash fallout simulation was carried out to contribute to the development of the hazard evaluation methodology. For the forest fire hazard, the concept of the hazard evaluation methodology was developed based on fire simulation. Event sequence assessment methodology was also developed based on plant dynamics analysis coupled with continuous Markov chain Monte Carlo method in order to apply to the event sequence against snow. In developing the strong wind PRA methodology, hazard curves were estimated by using Weibull and Gumbel distributions based on weather data recorded in Japan. The obtained hazard curves were divided into five discrete categories for event tree quantification. Next, failure probabilities for decay heat removal related components were calculated as a product of two probabilities: i.e., a probability for the missiles to enter the intake or out-take in the decay heat removal system, and fragility caused by the missile impacts. Finally, based on the event tree, the core damage frequency was estimated about 6*10 -9 /year by multiplying the discrete hazard probabilities in the Gumbel distribution by the conditional decay heat removal failure probabilities. A dominant sequence was led by the assumption that the operators could not extinguish fuel tank fire caused by the missile impacts and the fire induced loss of the decay heat removal system. (authors)

The radioprotective effect of a new aminothiol (20-PRA)

International Nuclear Information System (INIS)

Dolabela, M.F.; Lopes, M.T.P.; Pereira, M.T.; Steffani, G.M.; Pilo-Veloso, D.; Salas, C.E.; Nelson, D.L.

1998-01-01

We examined the radioprotective effect of aminothiol 2-N-propylamine-cyclohexane thiol (20-PRA) on a human leukemic cell line (K562) following various radiation doses (5,7.5 and 20 Gy) using a source of 60 Co γ-rays. At 5 Gy and 1nM 20-PRA, a substantial protective effect (58%) was seen 24 h after irradiation, followed by a decrease at 48 h (11%). At the high radiation dose (20 Gy) a low protective effect was also seen (35%). In addition, the anti tumorigenic potential of 10 nM 20-PRA was shown by the inhibition of crown gall formation induced by Agrobacterium tumefaciens. The radioprotective potency of 20-PRA is 10 5- 10 6 times higher than that of the aminothiol WR-1065 (N(2-mercaptoethyl)-1,3-diamino propane) whose protective effect is in the 0.1 to 1.0 nM range. (author)

The radioprotective effect of a new aminothiol (20-PRA

Directory of Open Access Journals (Sweden)

M.F. Dolabela

1998-08-01

Full Text Available We examined the radioprotective effect of aminothiol 2-N-propylamine-cyclo-hexanethiol (20-PRA on a human leukemic cell line (K562 following various radiation doses (5, 7.5 and 20 Gy using a source of 60Co g-rays. At 5 Gy and 1 nM 20-PRA, a substantial protective effect (58% was seen 24 h after irradiation, followed by a decrease at 48 h (11%. At the high radiation dose (20 Gy a low protective effect was also seen (35%. In addition, the antitumorigenic potential of 10 nM 20-PRA was shown by the inhibition of crown gall formation induced by Agrobacterium tumefaciens. The radioprotective potency of 20-PRA is 105-106 times higher than that of the aminothiol WR-1065 (N-(2-mercaptoethyl-1,3-diaminopropane whose protective effect is in the 0.1 to 1.0 mM range.

Clinical significance of determination of SAC/PRA value in patients with primary aldosteronism

International Nuclear Information System (INIS)

Li Liren; Dai Yaozong; Liu Jiumin

2003-01-01

Objective: To investigate the diagnostic significance of determining SAC/PRA valve in hyperaldosteronism. Methods: Plasma renin activity (PRA) and angiotensin (AT-II) as well as serum aldosterone contents were measured with RIA in 48 patients with primary aldosteronism and 30 controls. The SAC/PRA value was calculated. Results: Contents of PRA, AT-II and Aldo in blood of patients with primary aldosteronism were very significantly different from those in controls (p < 0.001) (PRA 0.14 ± 0.08 ng/ml/h vs 0.57 ± 0.08 ng/ml/h; AT-II 21.21 ± 7.55 ng/L vs 36.03 ± 6.11 ng/L; Aldo 1.07 ± 0.34 nmol/L vs 0.33 ± 0.04 nmol/L). Calculated SAC/PRA value was 913 ± 409 (normal upper limit 400). Conclusion: SAC/PRA value is an useful accessory diagnostic criterion for primary aldosteronism

Human factors assessment in PRA using Task Analysis Linked Evaluation Technique (TALENT)

International Nuclear Information System (INIS)

Wells, J.E.; Banks, W.W.

1991-01-01

Thirty years ago the US military and US aviation industry, and more recently, in response to the US Three Mile Island and USSR Chernobyl accidents, the US commercial nuclear power industry, acknowledged that human error, as an immediate precursor, and as a latent or indirect influence in the form of training, maintainability, inservice test, and surveillance programs, is a primary contributor to unreality and risk in complex high-reliability systems. A 1985 Nuclear Regulatory Commission (NRC) study of Licensee Event Reports (LERs) suggests that upwards of 65% of commercial nuclear system failures involve human error. Despite the magnitude and nature of human error cited in that study, there has been limited attention to personnel-centered issues, especially person-to-person issues involving group processes, management and organizational environment. The paper discusses NRC integration and applications research with respect to the Task Analysis Linked Evaluation Technique (TALENT) in risk assessment applications

Innovative probabilistic risk assessment applications: barrier impairments and fracture toughness. Panel Discussion

International Nuclear Information System (INIS)

Osterman, Michael; Root, Steven; Li, F.; Modarres, Mohammad; Reinhart, F. Mark; Bradley, Biff; Calhoun, David J.

2001-01-01

Full text of publication follows: New probabilistic risk assessment (PRA) applications promise to improve the overall safety and efficiency of nuclear plant operations. This discussion will explore the use of PRA in evaluating barrier integrity with respect to the consequences of natural phenomena such as tornadoes, floods, and harsh environments. Additionally, the session will explore proposals to improve fracture toughness techniques using PRA. (authors)

Application of FIVE methodology in probabilistic risk assessment (PRA) of fire events

International Nuclear Information System (INIS)

Lopez Garcia, F.J.; Suarez Alonso, J.; Fiolamengual, M.J.

1993-01-01

This paper reflects the experience acquired during the process of evaluation and updating of the fire analysis within the Cofrentes NPP PRA. It determines which points are the least precise, either because of their greater uncertainty or because of their excessive conservatism, as well as the subtasks which have involved a larger work load and could be simplified. These aspects are compared with the steps followed in methodology FIVE (Fire Vulnerability Evaluation Methodology) to assess whether application of this methodology would optimize the task, by making it more systematic and realistic and reducing uncertainties. On the one hand, the FIVE methodology does not have the scope sufficient to carry out a quantitative risk evaluation, but it can easily be complemented -without detriment to its systematic nature- by quantifying core damage in significant areas. On the other hand, certain issues such as definition of the fire growth software program which has to be used, are still not fully closed. Nevertheless, the conclusions derived from this assessment are satisfactory, since it is considered that this methodology would serve to unify the criteria and data of the analysis of fire-induced risks, providing a progressive screening method which would considerably simplify the task. (author)

Probabilistic Risk Assessment Process for High-Power Laser Operations in Outdoor Environments

Science.gov (United States)

2016-01-01

the NOHD/NSHD or its derivatives. Only those issued with an appropriate level of protection (such as protective eyewear or clothing) would be...developed in the emerging nuclear industry as well as in the established transport, petrochemical, and aerospace sectors, which were expanding...emergence of the PRA technique as an effective means of risk assessment. The origins of PRA lie in the aerospace industry .11,12 PRA is described by the

Integration of human reliability analysis into the probabilistic risk assessment process: Phase 1

International Nuclear Information System (INIS)

Bell, B.J.; Vickroy, S.C.

1984-10-01

A research program was initiated to develop a testable set of analytical procedures for integrating human reliability analysis (HRA) into the probabilistic risk assessment (PRA) process to more adequately assess the overall impact of human performance on risk. In this three-phase program, stand-alone HRA/PRA analytic procedures will be developed and field evaluated to provide improved methods, techniques, and models for applying quantitative and qualitative human error data which systematically integrate HRA principles, techniques, and analyses throughout the entire PRA process. Phase 1 of the program involved analysis of state-of-the-art PRAs to define the structures and processes currently in use in the industry. Phase 2 research will involve developing a new or revised PRA methodology which will enable more efficient regulation of the industry using quantitative or qualitative results of the PRA. Finally, Phase 3 will be to field test those procedures to assure that the results generated by the new methodologies will be usable and acceptable to the NRC. This paper briefly describes the first phase of the program and outlines the second

Organizational extension of PRA models and NASA application

International Nuclear Information System (INIS)

Pate-Cornell, E.

1989-01-01

This paper describes a probabilistic method which extends classical PRA to include some characteristics of the organization that processes or manages an engineering system. Ataxonomy of errors is presented and their organizational roots are examined. An assembly model is proposed for the analysis of the resulting spectrum of capacities of the system. The management of the Thermal Protection system of the Space Shuttle is used as an illustration. The model allows assessment of the benefits of organizational improvements of the orbiter's processing

Human reliability assessment and probabilistic risk assessment

International Nuclear Information System (INIS)

Embrey, D.E.; Lucas, D.A.

1989-01-01

Human reliability assessment (HRA) is used within Probabilistic Risk Assessment (PRA) to identify the human errors (both omission and commission) which have a significant effect on the overall safety of the system and to quantify the probability of their occurrence. There exist a variey of HRA techniques and the selection of an appropriate one is often difficult. This paper reviews a number of available HRA techniques and discusses their strengths and weaknesses. The techniques reviewed include: decompositional methods, time-reliability curves and systematic expert judgement techniques. (orig.)

Evaluation of allowed outage time using PRA results

International Nuclear Information System (INIS)

Johanson, G.

1985-01-01

In a probabilistic risk assessment (PRA) different measures of risk importance can be established. These measures can be used as a basis for further evaluation and determination of allowed outage time for specific components, within safety systems of a nuclear power plant. In order to optimize the allowed outage time (AOT) stipulated in the plant's Technical Specification it is necessary to create a methodology which could incorporate existing PRA data into a quantitative extrapolation. In order to evaluate the plant risk status due to AOT in a quantitative manner, the risk achievement worth is utilized. Risk achievement worth is defined as follows: to measure the worth of a feature, in achieving the present risk, one approach is to remove the feature and then determine how much the risk has increased. Thus, the risk achievement worth is formally defined to be the increase in risk if the feature were assumed not be there or to be failed. Another parameter of interest for this analysis is the shutdown risk increase. The shutdown risk achievement worth must be incorporated into the accident sequence risk achievement worth to arrive at an optimal set of plant specific AOTs

Uncertainty and sensitivity studies supporting the interpretation of the results of TVO I/II PRA

International Nuclear Information System (INIS)

Holmberg, J.

1992-01-01

A comprehensive Level 1 probabilistic risk assessment (PRA) has been performed for the TVO I/II nuclear power units. As a part of the PRA project, uncertainties of risk models and methods were systematically studied in order to describe them and to demonstrate their impact by way of results. The uncertainty study was divided into two phases: a qualitative and a quantitative study. The qualitative study contained identification of uncertainties and qualitative assessments of their importance. The PRA was introduced, and identified assumptions and uncertainties behind the models were documented. The most significant uncertainties were selected by importance measures or other judgements for further quantitative studies. The quantitative study included sensitivity studies and propagation of uncertainty ranges. In the sensitivity studies uncertain assumptions or parameters were varied in order to illustrate the sensitivity of the models. The propagation of the uncertainty ranges demonstrated the impact of the statistical uncertainties of the parameter values. The Monte Carlo method was used as a propagation method. The most significant uncertainties were those involved in modelling human interactions, dependences and common cause failures (CCFs), loss of coolant accident (LOCA) frequencies and pressure suppression. The qualitative mapping out of the uncertainty factors turned out to be useful in planning quantitative studies. It also served as internal review of the assumptions made in the PRA. The sensitivity studies were perhaps the most advantageous part of the quantitative study because they allowed individual analyses of the significance of uncertainty sources identified. The uncertainty study was found reasonable in systematically and critically assessing uncertainties in a risk analysis. The usefulness of this study depends on the decision maker (power company) since uncertainty studies are primarily carried out to support decision making when uncertainties are

Load out and offshore lifting of the PRA-1 platform modules; Embarque e icamento offshore dos modulos de PRA-1

Energy Technology Data Exchange (ETDEWEB)

Barbosa, Fernando; Raigorodsky, Jacques; Mitidieri, Jorge L.U.; Ricardi, Paulo S. [Construtora Norberto Odebrecht S.A., Rio de Janeiro, RJ (Brazil)

2008-07-01

The technology innovations are characteristics of offshore Engineering around the world. These technologies just make sense when they aim the productivity, security and costs gains compared to ordinary methods. It is in this context that the proposal of the Consorcio PRA-1 (Odebrecht e UTC) team makes sense, in the definition of basic methodology for the PRA-1 platform construction and installation. Through the innovative concept, It was defined (still in the proposal phase) the basic premise that the modules construction and assembly were onshore ending up that just few hours after the offshore installation the modules should be operational in minimal habitability conditions. This innovative method allowed the lack of Flotel, that is a platform which provide support to the offshore construction and assembly (Flotel represents a high costs to the project) and, as consequence, the contract signature by CONSORCIO PRA-1. This work aims to describe the method used for the LOUD-OUT of the PRA-1 modules and the installation of them on the jacket through a vessel provide with cranes the has performed the lifting. Theses operations became unique in Brazil due its challengers characteristics: Module 12 weight = 7203 tf and Module 35 = 5725 tf. For the accomplishment of the Load-out and offshore lifting, was performed a detailed planning and a high level of subcontract interface management. The operations mentioned above were filmed/photographed and published in the specialized media. (author)

Integration of human reliability analysis into the probabilistic risk assessment process: phase 1

International Nuclear Information System (INIS)

Bell, B.J.; Vickroy, S.C.

1985-01-01

The US Nuclear Regulatory Commission and Pacific Northwest Laboratory initiated a research program in 1984 to develop a testable set of analytical procedures for integrating human reliability analysis (HRA) into the probabilistic risk assessment (PRA) process to more adequately assess the overall impact of human performance on risk. In this three phase program, stand-alone HRA/PRA analytic procedures will be developed and field evaluated to provide improved methods, techniques, and models for applying quantitative and qualitative human error data which systematically integrate HRA principles, techniques, and analyses throughout the entire PRA process. Phase 1 of the program involved analysis of state-of-the-art PRAs to define the structures and processes currently in use in the industry. Phase 2 research will involve developing a new or revised PRA methodology which will enable more efficient regulation of the industry using quantitative or qualitative results of the PRA. Finally, Phase 3 will be to field test those procedures to assure that the results generated by the new methodologies will be usable and acceptable to the NRC. This paper briefly describes the first phase of the program and outlines the second

A probabilistic risk assessment of Oconee Unit 3. Executive highlights 60

International Nuclear Information System (INIS)

1984-04-01

In 1980 the Nuclear Safety Analysis Center and Duke Power Co. joined in a project to provide the utility industry with a practical, useful example of the application of probabilistic risk assessment (PRA) methods. PRA is a structured analysis technique that accounts for all the failure possibilities that might conceivably lead to core damage. The technique uses probabilities as discriminators to determine which are most significant. The following were project objectives: to provide the host utility with an analytic model of the plant that describes and estimates the likelihood of failure combinations that could lead to core melt; to evaluate the risks to the plant and to the public; to improve utility capabilities in PRA methods and applications

A desktop PRA

International Nuclear Information System (INIS)

Dolan, B.J.; Weber, B.J.

1989-01-01

This paper reports that Duke Power Company has completed full-scope PRAs for each of its nuclear stations - Oconee, McGuire and Catawba. These living PRAs are being maintained using desktop personal computers. Duke's PRA group now has powerful personal computer-based tools that have both decreased direct costs (computer analysis expenses) and increased group efficiency (less time to perform analyses). The shorter turnaround time has already resulted in direct savings through analyses provided in support of justification for continued station operation. Such savings are expected to continue with similar future support

Treatment of system dependencies and human interactions in PRA studies: a review and sensitivity study

International Nuclear Information System (INIS)

Orvis, D.D.; Joksimovich, V.; Worledge, D.H.

1985-01-01

The Electric Power Research Institute sponsored the review and comparison of five PRA studies: Arkansas Nuclear One - Unit 1, Big Rock Point, Grand Gulf, Limerick, and Zion - Unit 1. The review has been conducted in two phases. The Phase I review may be characterized as a qualitative look into many aspects of a PRA study. The Phase II review was performed to quantify the extent that differences in analytical techniques or key assumptions in these areas affect the differences in study results. In each of the PRA studies reviewed, the general descriptions of analytical approaches and descriptions of the analyses of event tree, fault tree and human interaction analyses that affected the dominant core damage sequences were reviewed. When these descriptions aroused interest because of seeming inconsistencies within the study or with other studies, they were pursued in some depth. The approaches or assumptions were contrasted to similar elements from other studies, and sensitivity analyses were performed in many cases to test the significance of results to the analytical models or assumptions. Inferences were drawn from the results regarding significance of the item to plant-specific results and, where possible, were generalized to other PRAs. This paper describes the results of the review of system dependencies and human interactions

Hiperurisemia pada Pra Diabetes

Directory of Open Access Journals (Sweden)

Ellyza Nasrul

2012-09-01

Full Text Available AbstrakAsam urat (AU merupakan produk akhir dari katabolisme adenin dan guanin yang berasal dari pemecahannukleotida purin. Urat dihasilkan oleh sel yang mengandung xanthine oxidase, terutama hepar dan usus kecil.Hiperurisemia adalah keadaan kadar asam urat dalam darah lebih dari 7,0 mg/dL.Pra diabetes adalah subjek yangmempunyai kadar glukosa plasma meningkat akan tetapi peningkatannya masih belum mencapai nilai minimaluntuk kriteria diagnosis diabetes melitus (DM. Glukosa darah puasa terganggu merupakan keadaan dimanapeningkatan kadar FPG≥100 mg/dL dan <126 mg/dL. Toleransi glukosa terganggu merupakan peningkatanglukosa plasma 2 jam setelah pembebanan 75 gram glukosa oral (≥140 mg/dL dan <200mg/dL dengan FPG<126 mg/dL.Insulin juga berperan dalam meningkatkan reabsorpsi asam urat di tubuli proksimal ginjal. Sehinggapada keadaan hiperinsulinemia pada pra diabetes terjadi peningkatan reabsorpsi yang akan menyebabkanhiperurisemia. Transporter urat yang berada di membran apikal tubuli renal dikenal sebagai URAT-1 berperandalam reabsorpsi urat.Kata kunci: Hiperurisemia, Pra DiabetesAbstractUric acid (AU is the end product of the catabolism of adenine and guanine nucleotides derived from thebreakdown of purines. Veins produced by cells containing xanthine oxidase, especially the liver and small intestine.Hyperuricemia is a state in the blood uric acid levels over 7.0 mg / dL.Pre-diabetes is a subject which has a plasmaglucose level will rise but the increase is still not reached the minimum value for the diagnostic criteria for diabetesmellitus (DM. Impaired fasting blood glucose is a condition in which increased levels of FPG ≥ 100 mg / dL and<126 mg / dL. Impaired glucose tolerance is an increase in plasma glucose 2 hours after 75 gram oral glucose load(≥ 140 mg / dL and <200mg/dl with FPG <126 mg / dL.Insulin also plays a role in increasing the reabsorption ofuric acid in renal proximal tubule. So that the hyperinsulinemia in the pre

Applications of Living Fire PRA models to Fire Protection Significance Determination Process in Taiwan

International Nuclear Information System (INIS)

De-Cheng, Chen; Chung-Kung, Lo; Tsu-Jen, Lin; Ching-Hui, Wu; Lin, James C.

2004-01-01

The living fire probabilistic risk assessment (PRA) models for all three operating nuclear power plants (NPPs) in Taiwan had been established in December 2000. In that study, a scenario-based PRA approach was adopted to systematically evaluate the fire and smoke hazards and associated risks. Using these fire PRA models developed, a risk-informed application project had also been completed in December 2002 for the evaluation of cable-tray fire-barrier wrapping exemption. This paper presents a new application of the fire PRA models to fire protection issues using the fire protection significance determination process (FP SDP). The fire protection issues studied may involve the selection of appropriate compensatory measures during the period when an automatic fire detection or suppression system in a safety-related fire zone becomes inoperable. The compensatory measure can either be a 24-hour fire watch or an hourly fire patrol. The living fire PRA models were used to estimate the increase in risk associated with the fire protection issue in terms of changes in core damage frequency (CDF) and large early release frequency (LERF). In compliance with SDP at-power and the acceptance guidelines specified in RG 1.174, the fire protection issues in question can be grouped into four categories; red, yellow, white and green, in accordance with the guidelines developed for FD SDP. A 24-hour fire watch is suggested only required for the yellow condition, while an hourly fire patrol may be adopted for the white condition. More limiting requirement is suggested for the red condition, but no special consideration is needed for the green condition. For the calculation of risk measures, risk impacts from any additional fire scenarios that may have been introduced, as well as more severe initiating events and fire damages that may accompany the fire protection issue should be considered carefully. Examples are presented in this paper to illustrate the evaluation process. (authors)

Computer aided probabilistic assessment of containment integrity

International Nuclear Information System (INIS)

Tsai, J.C.; Touchton, R.A.

1984-01-01

In the probabilistic risk assessment (PRA) of a nuclear power plant, there are three probability-based techniques which are widely used for event sequence frequency quantification (including nodal probability estimation). These three techniques are the event tree analysis, the fault tree analysis and the Bayesian approach for database development. In the barrier analysis for assessing radionuclide release to the environment in a PRA study, these techniques are employed to a greater extent in estimating conditions which could lead to failure of the fuel cladding and the reactor coolant system (RCS) pressure boundary, but to a lesser degree in the containment pressure boundary failure analysis. The main reason is that containment issues are currently still in a state of flux. In this paper, the authors describe briefly the computer programs currently used by the nuclear industry to do event tree analyses, fault tree analyses and the Bayesian update. The authors discuss how these computer aided probabilistic techniques might be adopted for failure analysis of the containment pressure boundary

Development of extreme rainfall PRA methodology for sodium-cooled fast reactor

International Nuclear Information System (INIS)

Nishino, Hiroyuki; Kurisaka, Kenichi; Yamano, Hidemasa

2016-01-01

The objective of this study is to develop a probabilistic risk assessment (PRA) methodology for extreme rainfall with focusing on decay heat removal system of a sodium-cooled fast reactor. For the extreme rainfall, annual excess probability depending on the hazard intensity was statistically estimated based on meteorological data. To identify core damage sequence, event trees were developed by assuming scenarios that structures, systems and components (SSCs) important to safety are flooded with rainwater coming into the buildings through gaps in the doors and the SSCs fail when the level of rainwater on the ground or on the roof of the building becomes higher than thresholds of doors on first floor or on the roof during the rainfall. To estimate the failure probability of the SSCs, the level of water rise was estimated by comparing the difference between precipitation and drainage capacity. By combining annual excess probability and the failure probability of SSCs, the event trees led to quantification of core damage frequency, and therefore the PRA methodology for rainfall was developed. (author)

Failure Modes Taxonomy for Reliability Assessment of Digital Instrumentation and Control Systems for Probabilistic Risk Analysis - Failure modes taxonomy for reliability assessment of digital I and C systems for PRA

International Nuclear Information System (INIS)

Amri, A.; Blundell, N.; ); Authen, S.; Betancourt, L.; Coyne, K.; Halverson, D.; Li, M.; Taylor, G.; Bjoerkman, K.; Brinkman, H.; Postma, W.; Bruneliere, H.; Chirila, M.; Gheorge, R.; Chu, L.; Yue, M.; Delache, J.; Georgescu, G.; Deleuze, G.; Quatrain, R.; Thuy, N.; Holmberg, J.-E.; Kim, M.C.; Kondo, K.; Mancini, F.; Piljugin, E.; Stiller, J.; Sedlak, J.; Smidts, C.; Sopira, V.

2015-01-01

Digital protection and control systems appear as upgrades in older nuclear power plants (NPP), and are commonplace in new NPPs. To assess the risk of NPP operation and to determine the risk impact of digital systems, there is a need to quantitatively assess the reliability of the digital systems in a justifiable manner. Due to the many unique attributes of digital systems (e.g., functions are implemented by software, units of the system interact in a communication network, faults can be identified and handled online), a number of modelling and data collection challenges exist, and international consensus on the reliability modelling has not yet been reached. The objective of the task group called DIGREL has been to develop a taxonomy of failure modes of digital components for the purposes of probabilistic risk analysis (PRA). An activity focused on the development of a common taxonomy of failure modes is seen as an important step towards standardised digital instrumentation and control (I and C) reliability assessment techniques for PRA. Needs from PRA has guided the work, meaning, e.g., that the I and C system and its failures are studied from the point of view of their functional significance point of view. The taxonomy will be the basis of future modelling and quantification efforts. It will also help to define a structure for data collection and to review PRA studies. The proposed failure modes taxonomy has been developed by first collecting examples of taxonomies provided by the task group organisations. This material showed some variety in the handling of I and C hardware failure modes, depending on the context where the failure modes have been defined. Regarding the software part of I and C, failure modes defined in NPP PRAs have been simple - typically a software CCF failing identical processing units. The DIGREL task group has defined a new failure modes taxonomy based on a hierarchical definition of five levels of abstraction: 1. system level (complete

Results and insights of a level-1 internal event PRA of a PWR during mid-loop operations

International Nuclear Information System (INIS)

Chu, T.L.; Musicki, Z.; Kohut, P.; Yang, J.; Bozoki, G.; Hsu, C.J.; Diamond, D.J.; Wong, S.M.; Holmes, B.; Su, R.F.; Dang, V.; Siu, N.; Bley, D.; Johnson, D.; Lin, J.

1994-01-01

Traditionally, probabilistic risk assessments (PRA) of severe accidents in nuclear power plants have considered initiating events potentially occurring only during full power operation. Some previous screening analysis that were performed for other modes of operation suggested that risks during those modes were small relative to full power operation. However, more recent studies and operational experience have implied that accidents during low power and shutdown could be significant contributors to risk. During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. The program includes two parallel projects being performed by BNL and SNL. Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The scope of the program includes that of a level-3 PRA. The objective of this paper is to present the approach utilized in the level-1 PRA for the Surry plant, and discuss the results obtained. A comparison of the results with those of other shutdown studies is provided. Relevant safety issues such as plant and hardware configurations, operator training, and instrumentation and control is discussed

The Prenylated Rab GTPase Receptor PRA1.F4 Contributes to Protein Exit from the Golgi Apparatus.

Science.gov (United States)

Lee, Myoung Hui; Yoo, Yun-Joo; Kim, Dae Heon; Hanh, Nguyen Hong; Kwon, Yun; Hwang, Inhwan

2017-07-01

Prenylated Rab acceptor1 (PRA1) functions in the recruitment of prenylated Rab proteins to their cognate organelles. Arabidopsis ( Arabidopsis thaliana ) contains a large number of proteins belonging to the AtPRA1 family. However, their physiological roles remain largely unknown. Here, we investigated the physiological role of AtPRA1.F4, a member of the AtPRA1 family. A T-DNA insertion knockdown mutant of AtPRA1.F4 , atpra1.f4 , was smaller in stature than parent plants and possessed shorter roots, whereas transgenic plants overexpressing HA:AtPRA1.F4 showed enhanced development of secondary roots and root hairs. However, both overexpression and knockdown plants exhibited increased sensitivity to high-salt stress, lower vacuolar Na + /K + -ATPase and plasma membrane ATPase activities, lower and higher pH in the vacuole and apoplast, respectively, and highly vesiculated Golgi apparatus. HA:AtPRA1.F4 localized to the Golgi apparatus and assembled into high-molecular-weight complexes. atpra1.f4 plants displayed a defect in vacuolar trafficking, which was complemented by low but not high levels of HA : AtPRA1.F4 Overexpression of HA:AtPRA1.F4 also inhibited protein trafficking at the Golgi apparatus, albeit differentially depending on the final destination or type of protein: trafficking of vacuolar proteins, plasma membrane proteins, and trans-Golgi network (TGN)-localized SYP61 was strongly inhibited; trafficking of TGN-localized SYP51 was slightly inhibited; and trafficking of secretory proteins and TGN-localized SYP41 was negligibly or not significantly inhibited. Based on these results, we propose that Golgi-localized AtPRA1.F4 is involved in the exit of many but not all types of post-Golgi proteins from the Golgi apparatus. Additionally, an appropriate level of AtPRA1.F4 is crucial for its function at the Golgi apparatus. © 2017 American Society of Plant Biologists. All Rights Reserved.

A framework to integrate software behavior into dynamic probabilistic risk assessment

International Nuclear Information System (INIS)

Zhu Dongfeng; Mosleh, Ali; Smidts, Carol

2007-01-01

Software plays an increasingly important role in modern safety-critical systems. Although, research has been done to integrate software into the classical probabilistic risk assessment (PRA) framework, current PRA practice overwhelmingly neglects the contribution of software to system risk. Dynamic probabilistic risk assessment (DPRA) is considered to be the next generation of PRA techniques. DPRA is a set of methods and techniques in which simulation models that represent the behavior of the elements of a system are exercised in order to identify risks and vulnerabilities of the system. The fact remains, however, that modeling software for use in the DPRA framework is also quite complex and very little has been done to address the question directly and comprehensively. This paper develops a methodology to integrate software contributions in the DPRA environment. The framework includes a software representation, and an approach to incorporate the software representation into the DPRA environment SimPRA. The software representation is based on multi-level objects and the paper also proposes a framework to simulate the multi-level objects in the simulation-based DPRA environment. This is a new methodology to address the state explosion problem in the DPRA environment. This study is the first systematic effort to integrate software risk contributions into DPRA environments

Reliability design of a critical facility: An application of PRA methods

International Nuclear Information System (INIS)

Souza Vieira Neto, A.; Souza Borges, W. de

1987-01-01

Although a general agreement concerning the enforcement of reliability (probabilistic) design criteria for nuclear utilities is yet to be achieved. PRA methodology can still be used successfully as a project design and review tool, aimed at improving system's prospective performance or minimizing expected accident consequences. In this paper, the potential of such an application of PRA methods is examined in the special case of a critical design project currently being developed in Brazil. (orig.)

40 CFR 180.1200 - Pseudomonas fluorescens strain PRA-25; temporary exemption from the requirement of a tolerance.

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 23 2010-07-01 2010-07-01 false Pseudomonas fluorescens strain PRA-25... RESIDUES IN FOOD Exemptions From Tolerances § 180.1200 Pseudomonas fluorescens strain PRA-25; temporary... established for residues of the microbial pesticide, pseudomonas fluorescens strain PRA-25 when used on peas...

System safety and reliability using object-oriented programming techniques

International Nuclear Information System (INIS)

Patterson-Hine, F.A.; Koen, B.V.

1987-01-01

Direct evaluation fault tree codes have been written in recursive, list-processing computer languages such as PL/1 (PATREC-I) and LISP (PATREC-L). The pattern-matching strategy implemented in these codes has been used extensively in France to evaluate system reliability. Recent reviews of the risk management process suggest that a data base containing plant-specific information be integrated with a package of codes used for probabilistic risk assessment (PRA) to alleviate some of the difficulties that make a PRA so costly and time-intensive. A new programming paradigm, object-oriented programming, is uniquely suited for the development of such a software system. A knowledge base and fault tree evaluation algorithm, based on previous experience with PATREC-L, have been implemented using object-oriented techniques, resulting in a reliability assessment environment that is easy to develop, modify, and extend

An evaluation of the reliability and usefulness of external-initiator PRA (probabilistic risk analysis) methodologies

Energy Technology Data Exchange (ETDEWEB)

Budnitz, R.J.; Lambert, H.E. (Future Resources Associates, Inc., Berkeley, CA (USA))

1990-01-01

The discipline of probabilistic risk analysis (PRA) has become so mature in recent years that it is now being used routinely to assist decision-making throughout the nuclear industry. This includes decision-making that affects design, construction, operation, maintenance, and regulation. Unfortunately, not all sub-areas within the larger discipline of PRA are equally mature,'' and therefore the many different types of engineering insights from PRA are not all equally reliable. 93 refs., 4 figs., 1 tab.

An evaluation of the reliability and usefulness of external-initiator PRA [probabilistic risk analysis] methodologies

International Nuclear Information System (INIS)

Budnitz, R.J.; Lambert, H.E.

1990-01-01

The discipline of probabilistic risk analysis (PRA) has become so mature in recent years that it is now being used routinely to assist decision-making throughout the nuclear industry. This includes decision-making that affects design, construction, operation, maintenance, and regulation. Unfortunately, not all sub-areas within the larger discipline of PRA are equally ''mature,'' and therefore the many different types of engineering insights from PRA are not all equally reliable. 93 refs., 4 figs., 1 tab

Uncertainty Quantification in the Reliability and Risk Assessment of Generation IV Reactors: Final Scientific/Technical Report

International Nuclear Information System (INIS)

Vierow, Karen; Aldemir, Tunc

2009-01-01

The project entitled, 'Uncertainty Quantification in the Reliability and Risk Assessment of Generation IV Reactors', was conducted as a DOE NERI project collaboration between Texas A and M University and The Ohio State University between March 2006 and June 2009. The overall goal of the proposed project was to develop practical approaches and tools by which dynamic reliability and risk assessment techniques can be used to augment the uncertainty quantification process in probabilistic risk assessment (PRA) methods and PRA applications for Generation IV reactors. This report is the Final Scientific/Technical Report summarizing the project.

Uncertainty Quantification in the Reliability and Risk Assessment of Generation IV Reactors: Final Scientific/Technical Report

Energy Technology Data Exchange (ETDEWEB)

Vierow, Karen; Aldemir, Tunc

2009-09-10

The project entitled, “Uncertainty Quantification in the Reliability and Risk Assessment of Generation IV Reactors”, was conducted as a DOE NERI project collaboration between Texas A&M University and The Ohio State University between March 2006 and June 2009. The overall goal of the proposed project was to develop practical approaches and tools by which dynamic reliability and risk assessment techniques can be used to augment the uncertainty quantification process in probabilistic risk assessment (PRA) methods and PRA applications for Generation IV reactors. This report is the Final Scientific/Technical Report summarizing the project.

Mutation of praR in Rhizobium leguminosarum enhances root biofilms, improving nodulation competitiveness by increased expression of attachment proteins.

Science.gov (United States)

Frederix, Marijke; Edwards, Anne; Swiderska, Anna; Stanger, Andrew; Karunakaran, Ramakrishnan; Williams, Alan; Abbruscato, Pamela; Sanchez-Contreras, Maria; Poole, Philip S; Downie, J Allan

2014-08-01

In Rhizobium leguminosarum bv. viciae, quorum-sensing is regulated by CinR, which induces the cinIS operon. CinI synthesizes an AHL, whereas CinS inactivates PraR, a repressor. Mutation of praR enhanced biofilms in vitro. We developed a light (lux)-dependent assay of rhizobial attachment to roots and demonstrated that mutation of praR increased biofilms on pea roots. The praR mutant out-competed wild-type for infection of pea nodules in mixed inoculations. Analysis of gene expression by microarrays and promoter fusions revealed that PraR represses its own transcription and mutation of praR increased expression of several genes including those encoding secreted proteins (the adhesins RapA2, RapB and RapC, two cadherins and the glycanase PlyB), the polysaccharide regulator RosR, and another protein similar to PraR. PraR bound to the promoters of several of these genes indicating direct repression. Mutations in rapA2, rapB, rapC, plyB, the cadherins or rosR did not affect the enhanced root attachment or nodule competitiveness of the praR mutant. However combinations of mutations in rapA, rapB and rapC abolished the enhanced attachment and nodule competitiveness. We conclude that relief of PraR-mediated repression determines a lifestyle switch allowing the expression of genes that are important for biofilm formation on roots and the subsequent initiation of infection of legume roots. © 2014 The Authors. Molecular Microbiology published by John Wiley & Sons Ltd.

Applications of PRA in nuclear criticality safety

International Nuclear Information System (INIS)

McLaughlin, T.P.

1992-01-01

Traditionally, criticality accident prevention at Los Alamos has been based on a thorough review and understanding of proposed operations of changes to operations, involving both process supervision and criticality safety staff. The outcome of this communication was usually an agreement, based on professional judgement, that certain accident sequences were credible and had to be reduced in likelihood either by administrative controls or by equipment design and others were not credible, and thus did not warrant expenditures to further reduce their likelihood. The extent of analysis and documentation was generally in proportion to the complexity of the operation but did not include quantified risk assessments. During the last three years nuclear criticality safety related Probabilistic Risk Assessments (PRAs) have been preformed on operations in two Los Alamos facilities. Both of these were conducted in order to better understand the cost/benefit aspects of PRA's as they apply to largely ''hands-on'' operations with fissile material for which human errors or equipment failures significant to criticality safety are both rare and unique. Based on these two applications and an appreciation of the historical criticality accident record (frequency and consequences) it is apparent that quantified risk assessments should be performed very selectively

Results and insights of a level-1 internal event PRA of a PWR during mid-loop operations

International Nuclear Information System (INIS)

Chu, T.L.; Musicki, Z.; Kohut, P.

1993-01-01

Traditionally, probabilistic risk assessments (PRA) of severe accidents in nuclear power plants have considered initiating events potentially occurring only during full power operation. Some previous screening analysis that were performed for other modes of operation suggested that risks during those modes were small relative to full power operation. However, more recent studies and operational experience have implied that accidents during low power and shutdown could be significant contributors to risk. The objective of this paper is to present the approach utilized in the level-1 PRA for the Surry plant, and discuss the results obtained. A comparison of the results with those of other shutdown studies is provided. Relevant safety issues such as plant and hardware configurations, operator training, and instrumentation and control is discussed

Current status and future expectation concerning probabilistic risk assessment of NPPs. 1. Features and issues of probabilistic risk assessment methodology

International Nuclear Information System (INIS)

Yamashita, Masahiro

2012-01-01

Probabilistic risk assessment (PRA) of Nuclear Power Plants (NPPs) could play an important role in assuring safety of NPPs. However PRA had not always effectively used, which was indicated in Japanese government's report on Fukushima Daiichi NPP accident. At the Risk Technical Committee (RTC) of Standards Committee of Atomic Energy Society of Japan, preparation of standards (implementing criteria) focusing on PRA methodology and investigation on basic philosophy for use of PRA had been in progress. Based on activities of RTC, a serial in three articles including this described current status and future expectation concerning probabilistic risk assessment of NPPs. This article introduced features and issues of PRA methodology related to the use of PRA. Features of PRA methodology could be shown as (1) systematic and comprehensive understanding of risk, (2) support of grading approach, (3) identification of effective safety upgrade measures and (4) quantitative understanding of effects of uncertainty. Issues of PRA methodology were (1) extension of PRA application area, (2) upgrade of PRA methodology, (3) quality assurance of PRA, (4) treatment of uncertainty and (5) quantitative evaluation criteria. (T. Tanaka)

Manutenção de brinquedo em praças públicas

Directory of Open Access Journals (Sweden)

Fabio Namiki

2007-12-01

Full Text Available O artigo apresenta o jacaré, um dos brinquedos executados no âmbito do Programa Centros de Bairro, que foi responsável pela implantação de cerca de 50 praças na cidade de São Paulo entre 2002 e 2004. O conjunto dos brinquedos deste programa foi apresentado e analisado no mestrado “Manutenção de praças na cidade de São Paulo. Estudo de caso: brinquedos do programa Centros de Bairro”, segundo metodologia que pode ser também aplicada para outros componentes de uma praça e mesmo para a praça em si. Espera-se que esta metodologia sirva como instrumento para o planejamento das ações de manutenção de praças e de mobiliários urbanos de modo geral. Neste texto, são apresentadas informações (da mesma forma que seriam em um manual de uso, operação e manutenção do projeto do brinquedo, obtidas junto aos responsáveis pelo programa, em entrevista com o executor dos brinquedos e através dos desenhos e documentos produzidos para a licitação e execução das peças. São também apresentadas as informações obtidas a partir das inspeções a campo e estimativas do custo de manutenção preventiva. Frente ao custo de reposição de um brinquedo novo, os valores da manutenção nos provam a importância econômica de tais ações.

Probabilistic risk assessment and its role in plant modifications

International Nuclear Information System (INIS)

Diederich, A.R.; McElroy, W.F.

1986-01-01

Electric Utilities today have a tool available to improve management's ability to evaluate nuclear power plant modifications (MODS). Probabilistic Risk Assessment (PRA), is a tool of choice since it can be applied to a specific situation such as MOD request review, bringing the perspectives of reliability, financial risk and consequences to the public in addition to the more rigid requirements like those associated with Quality Assurance or licensing criteria. The techniques used in the PRA process revolve about the creation and manipulation of Fault Trees and Event Trees, which are used to quantify the event sequences and reliability of plant systems in a logical framework. It is through these methods that chains of sequences, or events, are understood. The degree to which plant systems are modelled in the PRA can vary depending on resources and purpose. Philadelphia Elecrtric Company's PRA modelled ten (10) major systems but this number may increase during the application and updating process

Implications of an HRA framework for quantifying human acts of commission and dependency: Development of a methodology for conducting an integrated HRA/PRA

International Nuclear Information System (INIS)

Barriere, M.T.; Luckas, W.J.; Brown, W.S.; Cooper, S.E.; Wreathall, J.; Bley, D.C.

1994-01-01

To support the development of a refined human reliability analysis (HRA) framework, to address identified HRA user needs and improve HRA modeling, unique aspects of human performance have been identified from an analysis of actual plant-specific events. Through the use of the refined framework, relationships between the following HRA, human factors and probabilistic risk assessment (PRA) elements were described: the PRA model, plant states, plant conditions, PRA basic events, unsafe human actions, error mechanisms, and performance shaping factors (PSFs). The event analyses performed in the context of the refined HRA framework, identified the need for new HRA methods that are capable of: evaluating a range of different error mechanisms (e.g., slips as well as mistakes); addressing errors of commission (EOCs) and dependencies between human actions; and incorporating the influence of plant conditions and multiple PSFs on human actions. This report discusses the results of the assessment of user needs, the refinement of the existing HRA framework, as well as, the current status on EOCs, and human dependencies

Implications of an HRA framework for quantifying human acts of commission and dependency: Development of a methodology for conducting an integrated HRA/PRA

International Nuclear Information System (INIS)

Barriere, M.T.; Luckas, W.J.; Brown, W.S.; Cooper, S.E.; Wreathall, J.; Bley, D.C.

1993-01-01

To support the development of a refined human reliability analysis (HRA) framework, to address identified HRA user needs and improve HRA modeling, unique aspects of human performance have been identified from an analysis of actual plant-specific events. Through the use of the refined framework, relationships between the following HRA, human factors and probabilistic risk assessment (PRA) elements were described: the PRA model, plant states, plant conditions, PRA basic events, unsafe human actions, error mechanisms, and performance shaping factors (PSFs). The event analyses performed in the context of the refined HRA framework, identified the need for new HRA methods that are capable of: evaluating a range of different error mechanisms (e.g., slips as well as mistakes); addressing errors of commission (EOCs) and dependencies between human actions; and incorporating the influence of plant conditions and multiple PSFs on human actions. This report discusses the results of the assessment of user needs, the refinement of the existing HRA framework, as well as, the current status on EOCs, and human dependencies

Analytical solutions of linked fault tree probabilistic risk assessments using binary decision diagrams with emphasis on nuclear safety applications

International Nuclear Information System (INIS)

Nusbaumer, O. P. M.

2007-01-01

This study is concerned with the quantification of Probabilistic Risk Assessment (PRA) using linked Fault Tree (FT) models. Probabilistic Risk assessment (PRA) of Nuclear Power Plants (NPPs) complements traditional deterministic analysis; it is widely recognized as a comprehensive and structured approach to identify accident scenarios and to derive numerical estimates of the associated risk levels. PRA models as found in the nuclear industry have evolved rapidly. Increasingly, they have been broadly applied to support numerous applications on various operational and regulatory matters. Regulatory bodies in many countries require that a PRA be performed for licensing purposes. PRA has reached the point where it can considerably influence the design and operation of nuclear power plants. However, most of the tools available for quantifying large PRA models are unable to produce analytically correct results. The algorithms of such quantifiers are designed to neglect sequences when their likelihood decreases below a predefined cutoff limit. In addition, the rare event approximation (e.g. Moivre's equation) is typically implemented for the first order, ignoring the success paths and the possibility that two or more events can occur simultaneously. This is only justified in assessments where the probabilities of the basic events are low. When the events in question are failures, the first order rare event approximation is always conservative, resulting in wrong interpretation of risk importance measures. Advanced NPP PRA models typically include human errors, common cause failure groups, seismic and phenomenological basic events, where the failure probabilities may approach unity, leading to questionable results. It is accepted that current quantification tools have reached their limits, and that new quantification techniques should be investigated. A novel approach using the mathematical concept of Binary Decision Diagram (BDD) is proposed to overcome these deficiencies

Seismic margin analysis technique for nuclear power plant structures

International Nuclear Information System (INIS)

Seo, Jeong Moon; Choi, In Kil

2001-04-01

In general, the Seismic Probabilistic Risk Assessment (SPRA) and the Seismic Margin Assessment(SAM) are used for the evaluation of realistic seismic capacity of nuclear power plant structures. Seismic PRA is a systematic process to evaluate the seismic safety of nuclear power plant. In our country, SPRA has been used to perform the probabilistic safety assessment for the earthquake event. SMA is a simple and cost effective manner to quantify the seismic margin of individual structural elements. This study was performed to improve the reliability of SMA results and to confirm the assessment procedure. To achieve this goal, review for the current status of the techniques and procedures was performed. Two methodologies, CDFM (Conservative Deterministic Failure Margin) sponsored by NRC and FA (Fragility Analysis) sponsored by EPRI, were developed for the seismic margin review of NPP structures. FA method was originally developed for Seismic PRA. CDFM approach is more amenable to use by experienced design engineers including utility staff design engineers. In this study, detailed review on the procedures of CDFM and FA methodology was performed

Methodology and application of surrogate plant PRA analysis to the Rancho Seco Power Plant: Final report

International Nuclear Information System (INIS)

Gore, B.F.; Huenefeld, J.C.

1987-07-01

This report presents the development and the first application of generic probabilistic risk assessment (PRA) information for identifying systems and components important to public risk at nuclear power plants lacking plant-specific PRAs. A methodology is presented for using the results of PRAs for similar (surrogate) plants, along with plant-specific information about the plant of interest and the surrogate plants, to infer important failure modes for systems of the plant of interest. This methodology, and the rationale on which it is based, is presented in the context of its application to the Rancho Seco plant. The Rancho Seco plant has been analyzed using PRA information from two surrogate plants. This analysis has been used to guide development of considerable plant-specific information about Rancho Seco systems and components important to minimizing public risk, which is also presented herein

Real-time risk assessment of operational events

International Nuclear Information System (INIS)

Perryman, L.J.; Foster, N.A.S.; Nicholls, D.R.; Grobbelaar, J.F.

1996-01-01

Probabilistic risk assessment (PRA) has always been fundamental to the licensing process of Koeberg nuclear power station. Furthermore, over the past 8 years PRA has assisted in many areas of operation. One of these areas is the real-time assessment of abnormal operating events. Over the years, considerable experience has been gained in using PRA to improve plant safety and performance. This paper presents some of the insights obtained in using PRA in such a dynamic role and demonstrates that, by developing and using the plant-specific 'living' PRA, considerable safety and financial gains can be obtained. These insights specifically concern the prerequisites before optimal use of a plant-specific 'living' PRA can be made. Finally, examples are presented of occurrences when PRA was used to improve plant safety and performance. These examples serve to demonstrate the advantages that can be obtained if sufficient resources are placed at the disposal of the PRA team. (orig.)

Safety analysis, risk assessment, and risk acceptance criteria

International Nuclear Information System (INIS)

Jamali, K.

1997-01-01

This paper discusses a number of topics that relate safety analysis as documented in the Department of Energy (DOE) safety analysis reports (SARs), probabilistic risk assessments (PRA) as characterized primarily in the context of the techniques that have assumed some level of formality in commercial nuclear power plant applications, and risk acceptance criteria as an outgrowth of PRA applications. DOE SARs of interest are those that are prepared for DOE facilities under DOE Order 5480.23 and the implementing guidance in DOE STD-3009-94. It must be noted that the primary area of application for DOE STD-3009 is existing DOE facilities and that certain modifications of the STD-3009 approach are necessary in SARs for new facilities. Moreover, it is the hazard analysis (HA) and accident analysis (AA) portions of these SARs that are relevant to the present discussions. Although PRAs can be qualitative in nature, PRA as used in this paper refers more generally to all quantitative risk assessments and their underlying methods. HA as used in this paper refers more generally to all qualitative risk assessments and their underlying methods that have been in use in hazardous facilities other than nuclear power plants. This discussion includes both quantitative and qualitative risk assessment methods. PRA has been used, improved, developed, and refined since the Reactor Safety Study (WASH-1400) was published in 1975 by the Nuclear Regulatory Commission (NRC). Much debate has ensued since WASH-1400 on exactly what the role of PRA should be in plant design, reactor licensing, 'ensuring' plant and process safety, and a large number of other decisions that must be made for potentially hazardous activities. Of particular interest in this area is whether the risks quantified using PRA should be compared with numerical risk acceptance criteria (RACs) to determine whether a facility is 'safe.' Use of RACs requires quantitative estimates of consequence frequency and magnitude

Two decades of PRA: What next?

International Nuclear Information System (INIS)

Rasmussen, N.C.

1992-01-01

Two decades ago, in the spring of 1972, the Reactor Safety Study was undertaken for the US Atomic Energy Commission (AEC). The goal of this study was to assess the risk to the public posed by the nuclear power plants operating in the US. Some three and one-half years later in October 1975, the study group issued its final report titled The Reactor Safety Study, also commonly known by its document number WASH 1400. Because it was issued at a time of heated public debate about nuclear safety, WASH 1400 received considerable critical review. By the late 1970s, as a result of the Lewis Report and the accident at Three Mile Island, the value of the WASH 1400 methodology was gradually recognized. A number of utilities undertook such studies of their own plants. The field of probabilistic risk assessment (PRA) developed from these efforts. Challenges remain. Among these are how to effectively communicate the results of the analysis. Just what does a probability of one in a million mean? Is there a de minimis probability - one so small that it can be ignored? How should society make decisions under substantial uncertainty? A number of these questions pose real challenges for the future

Analytical solutions of linked fault tree probabilistic risk assessments using binary decision diagrams with emphasis on nuclear safety applications[Dissertation 17286

Energy Technology Data Exchange (ETDEWEB)

Nusbaumer, O. P. M

2007-07-01

This study is concerned with the quantification of Probabilistic Risk Assessment (PRA) using linked Fault Tree (FT) models. Probabilistic Risk assessment (PRA) of Nuclear Power Plants (NPPs) complements traditional deterministic analysis; it is widely recognized as a comprehensive and structured approach to identify accident scenarios and to derive numerical estimates of the associated risk levels. PRA models as found in the nuclear industry have evolved rapidly. Increasingly, they have been broadly applied to support numerous applications on various operational and regulatory matters. Regulatory bodies in many countries require that a PRA be performed for licensing purposes. PRA has reached the point where it can considerably influence the design and operation of nuclear power plants. However, most of the tools available for quantifying large PRA models are unable to produce analytically correct results. The algorithms of such quantifiers are designed to neglect sequences when their likelihood decreases below a predefined cutoff limit. In addition, the rare event approximation (e.g. Moivre's equation) is typically implemented for the first order, ignoring the success paths and the possibility that two or more events can occur simultaneously. This is only justified in assessments where the probabilities of the basic events are low. When the events in question are failures, the first order rare event approximation is always conservative, resulting in wrong interpretation of risk importance measures. Advanced NPP PRA models typically include human errors, common cause failure groups, seismic and phenomenological basic events, where the failure probabilities may approach unity, leading to questionable results. It is accepted that current quantification tools have reached their limits, and that new quantification techniques should be investigated. A novel approach using the mathematical concept of Binary Decision Diagram (BDD) is proposed to overcome these

Insights into PRA methodologies

International Nuclear Information System (INIS)

Gallagher, D.; Lofgren, E.; Atefi, B.; Liner, R.; Blond, R.; Amico, P.

1984-08-01

Probabilistic Risk Assessments (PRAs) for six nuclear power plants were examined to gain insight into how the choice of analytical methods can affect the results of PRAs. The PRA sreflectope considered was limited to internally initiated accidents sequences through core melt. For twenty methodological topic areas, a baseline or minimal methodology was specified. The choice of methods for each topic in the six PRAs was characterized in terms of the incremental level of effort above the baseline. A higher level of effort generally reflects a higher level of detail or a higher degree of sophistication in the analytical approach to a particular topic area. The impact on results was measured in terms of how additional effort beyond the baseline level changed the relative importance and ordering of dominant accident sequences compared to what would have been observed had methods corresponding to the baseline level of effort been employed. This measure of impact is a more useful indicator of how methods affect perceptions of plant vulnerabilities than changes in core melt frequency would be. However, the change in core melt frequency was used as a secondary measure of impact for nine topics where availability of information permitted. Results are presented primarily in the form of effort-impact matrices for each of the twenty topic areas. A suggested effort-impact profile for future PRAs is presented

Probabilistic risk assessment in the CPI

International Nuclear Information System (INIS)

Guymer, P.; Kaiser, G.D.; Mc Kelvey, T.C.; Hannaman, G.W.

1987-01-01

Probabilistic Risk Assessment (PRA) is a method of quantifying the frequency of occurrence and magnitude of the consequences of accidents in systems that contain hazardous materials such as toxic, flammable or explosive chemicals. The frequency and magnitude of the consequences are the basic elements in the definition of risk, often simply expressed as the product of frequency and magnitude, summed over all accident sequences. PRA is a mature technique that has been used to estimate risk for a number of industrial facilities: for example, the Canvey Island Petrochemical complex; the Port of Rotterdam; the Reactor Safety Study, the first study to put the risks associated with nuclear power into perspective; and the transportation of chlorine. PRA has been developed to a greater level of sophistication in the nuclear industry than in the chemical industry. In the nuclear area, its usefulness has been demonstrated by increased plant safety, engineering insights, and cost-saving recommendations. Data and methods have been developed to increase the level of realism of the treatment of operator actions in PRA studies. It can be stated generally that the same methods can be applied with equal success in the chemical industry. However, there are pitfalls into which the unwary nuclear-oriented PRA analyst may stumble if he does not bear in mind that there are significant differences between nuclear plants and chemical plants

Pemikiran Suksesi Dalam Politik Islam Masa Pra Modern

Directory of Open Access Journals (Sweden)

Mazro'atus Sa'adah

2016-12-01

Abstrak: Pemikiran politik Islam muncul setelah Islam melalui Nabi Muhammad SAW berhasil membentuk sebuah ummat baru, dari peralihan kekuasaan kerajaan/kesukuan kepada Nabi yang kemudian kepada umat. Nabi Muhammad dinilai berhasil dalam mengatur komunitas barunya yang dikendalikan oleh ajarannya dalam seluruh lini kehidupan. Persoalan muncul kemudian setelah beliau wafat, yang akhirnya memunculkan pemikiran tentang suksesi. Artikel ini akan membahas tentang mengapa terjadi suksesi setelah Nabi Muhammad SAW wafat, bagaimana pemikiran para tokoh politik Islam masa pra modern terkait dengan suksesi, dan apa kontribusi pemikiran suksesi ini terhadap politik Islam di Indonesia. Dengan menggunakan pendekatan sejarah, ditemukan bahwa Nabi Muhammad tidak menetapkan siapa yang akan menggantikannya, dan ketika beliau wafat (632 M, para sahabat memilih seorang pemimpin (imam/khalifah. Masa pemerintahan Abu Bakar, Umar dan Usman banyak terjadi perselisihan yang awalnya terkait kepentingan agama namun berkembang menjadi kepentingan politik. Ketika Ali bin Abi Talib diangkat sebagai khalifah, konflik politik berkepanjangan berkaitan dengan pembunuhan Usman, menjadikan timbulnya perang jamal antara Aisyah dan Ali. Pada masa ini perbedaan kepentingan aqidah dipolitisir lebih jauh menjadi sebuah kepentingan politik. Dinamika politik ini kemudian melahirkan mazhab politik Islam klasik yang terbagi dalam tiga mazhab besar yaitu Sunni, Syi'ah dan Khawarij, yang darinya muncul istilah-istilah khilafah, imamah, ahlul halli wal aqdi, bay’ah, walayah dan lain-lain. Dari ketiga mazhab politik ini, kemudian muncul ide pemikiran politik Islam yang sangat kompleks dan berkepanjangan dari para tokoh politik Islam pra modern yang banyak dipengaruhi oleh filosof Yunani. Di Indonesia, pemikiran suksesi dalam politik Islam masa pra modern ini pernah diwacanakan. Namun untuk pemilihan kepala Negara belum terealisasi mengingat Indonesia bukan Negara Islam.

ANALYSIS OF AGRO-ECOLOGICAL SITUATION FOR IDENTIFICATION OF PROBLEMS BY PRA TECHNIQUES IN ADAPTIVE VILLAGE OF KRISHI VIGYAN KENDRA UNDER NEW ALLUVIA ZONE OF MURSHIDABAD DISTRICT OF WEST BENGAL

Directory of Open Access Journals (Sweden)

Abhishake Naskar

2011-07-01

Full Text Available Agro Ecosystem analysis using the Participatory Rural Appraisal (PRA techniques of an adaptive village (Jainpur of New Alluvial Zone of Murshidabad-Jiaganj block in Murshidabad district, West Bengal revealed that the village basically has rice and jute based farming system. The cropping intensity of the village is 233%. Out of 363 household 80% is engaged in Agriculture, Animal Husbandry and other allied activities. The land availability per household is 0.40 ha. The villagers are mostly scheduled caste. By snow ball technique major problems were identified .On the basis of bio-physical and socio-economic problems, thrust area were selected. Area specific On Farm Trials (OFT in farmers' were conducted on some researchable issues. Front Line Demonstration (FLD, training programme, health camp, awareness camp and other different extension activities were arranged to mitigate the problems.

Urgensi Pemeriksaan Psikis Pra-Nikah (Studi Pandangan Kepala KUA dan Psikolog Kota Malang

Directory of Open Access Journals (Sweden)

Ika Kurnia Fitriani

2015-06-01

Full Text Available Beberapa negara muslim memberikan perhatian terhadap pemeriksaan psikis pra-nikah bagi calon mempelai, sebagai upaya menanggulangi masalah rumah tangga akibat gangguan kejiwaan di masa yang akan datang. Penelitian ini bertujuan menggali informasi dari Kepala KUA dan Psikolog di Kota Malang tentang pemeriksaan psikis pra-nikah dan urgensinya bagi calon mempelai. Penelitian ini termasuk dalam penelitian lapangan (field reasearch, dengan menggunakan pendekatan kualitatif.  Alanisis data dilakukan melalui tiga tahapan yaitu reduksi data, penyajian data, dan menarik kesimpulan. Pengecekan keabsahan data menggunakan triangulasi sumber yang membandingkan hasil wawancara dengan data sekunder, dan triangulasi teori. Hasil dari penelitian ini menunjukkan bahwa Kepala KUA dan Psikolog di kota Malang menyetujui diadakan pemeriksaan psikis pranikah akan tetapi harus ada aturan hukumnya dan dilakukan sosialisasi agar program menjadi efektif. Selain itu, pemeriksaan psikis pra-nikah tidak bertentangan dengan konsep maqashid al-syari’ah dan konsep sadz al-dzari’ah dalam hukum Islam.

Estrogen and progesterone receptors have distinct roles in the establishment of the hyperplastic phenotype in PR-A transgenic mice

Energy Technology Data Exchange (ETDEWEB)

Simian, Marina; Bissell, Mina J.; Barcellos-Hoff, Mary Helen; Shyamala, Gopalan

2009-05-11

Expression of the A and B forms of progesterone receptor (PR) in an appropriate ratio is critical for mammary development. Mammary glands of PR-A transgenic mice, carrying an additional A form of PR as a transgene, exhibit morphological features associated with the development of mammary tumors. Our objective was to determine the roles of estrogen (E) and progesterone (P) in the genesis of mammary hyperplasias/preneoplasias in PR-A transgenics. We subjected PR-A mice to hormonal treatments and analyzed mammary glands for the presence of hyperplasias and used BrdU incorporation to measure proliferation. Quantitative image analysis was carried out to compare levels of latency-associated peptide and transforming growth factor beta 1 (TGF{beta}1) between PR-A and PR-B transgenics. Basement membrane disruption was examined by immunofluorescence and proteolytic activity by zymography. The hyperplastic phenotype of PR-A transgenics is inhibited by ovariectomy, and is reversed by treatment with E + P. Studies using the antiestrogen ICI 182,780 or antiprogestins RU486 or ZK 98,299 show that the increase in proliferation requires signaling through E/estrogen receptor alpha but is not sufficient to give rise to hyperplasias, whereas signaling through P/PR has little impact on proliferation but is essential for the manifestation of hyperplasias. Increased proliferation is correlated with decreased TGF{beta}1 activation in the PR-A transgenics. Analysis of basement membrane integrity showed loss of laminin-5, collagen III and collagen IV in mammary glands of PR-A mice, which is restored by ovariectomy. Examination of matrix metalloproteases (MMPs) showed that total levels of MMP-2 correlate with the steady-state levels of PR, and that areas of laminin-5 loss coincide with those of activation of MMP-2 in PR-A transgenics. Activation of MMP-2 is dependent on treatment with E and P in ovariectomized wild-type mice, but is achieved only by treatment with P in PR-A mice. These data

Probabilistic risk assessment course documentation. Volume 4. System reliability and analysis techniques sessions B/C - event trees/fault trees

International Nuclear Information System (INIS)

Haasl, D.; Young, J.

1985-08-01

This course will employ a combination of lecture material and practical problem solving in order to develop competence and understanding of th principles and techniques of event tree and fault tree analysis. The role of these techniques in the overall context of PRA will be described. The emphasis of this course will be on the basic, traditional methods of event tree and fault tree analysis

RAVEN: a GUI and an Artificial Intelligence Engine in a Dynamic PRA Framework

Energy Technology Data Exchange (ETDEWEB)

C. Rabiti; D. Mandelli; A. Alfonsi; J. Cogliati; R. Kinoshita; D. Gaston; R. Martineau; C. Curtis

2013-06-01

Increases in computational power and pressure for more accurate simulations and estimations of accident scenario consequences are driving the need for Dynamic Probabilistic Risk Assessment (PRA) [1] of very complex models. While more sophisticated algorithms and computational power address the back end of this challenge, the front end is still handled by engineers that need to extract meaningful information from the large amount of data and build these complex models. Compounding this problem is the difficulty in knowledge transfer and retention, and the increasing speed of software development. The above-described issues would have negatively impacted deployment of the new high fidelity plant simulator RELAP-7 (Reactor Excursion and Leak Analysis Program) at Idaho National Laboratory. Therefore, RAVEN that was initially focused to be the plant controller for RELAP-7 will help mitigate future RELAP-7 software engineering risks. In order to accomplish this task, Reactor Analysis and Virtual Control Environment (RAVEN) has been designed to provide an easy to use Graphical User Interface (GUI) for building plant models and to leverage artificial intelligence algorithms in order to reduce computational time, improve results, and help the user to identify the behavioral pattern of the Nuclear Power Plants (NPPs). In this paper we will present the GUI implementation and its current capability status. We will also introduce the support vector machine algorithms and show our evaluation of their potentiality in increasing the accuracy and reducing the computational costs of PRA analysis. In this evaluation we will refer to preliminary studies performed under the Risk Informed Safety Margins Characterization (RISMC) project of the Light Water Reactors Sustainability (LWRS) campaign [3]. RISMC simulation needs and algorithm testing are currently used as a guidance to prioritize RAVEN developments relevant to PRA.

Application of the NUREG/CR-6850 EPRI/NRC Fire PRA Methodology to a DOE Facility

International Nuclear Information System (INIS)

Elicson, Tom; Harwood, Bentley; Yorg, Richard; Lucek, Heather; Bouchard, Jim; Jukkola, Ray; Phan, Duan

2011-01-01

The application NUREG/CR-6850 EPRI/NRC fire PRA methodology to DOE facility presented several challenges. This paper documents the process and discusses several insights gained during development of the fire PRA. A brief review of the tasks performed is provided with particular focus on the following: Tasks 5 and 14: Fire-induced risk model and fire risk quantification. A key lesson learned was to begin model development and quantification as early as possible in the project using screening values and simplified modeling if necessary. Tasks 3 and 9: Fire PRA cable selection and detailed circuit failure analysis. In retrospect, it would have been beneficial to perform the model development and quantification in 2 phases with detailed circuit analysis applied during phase 2. This would have allowed for development of a robust model and quantification earlier in the project and would have provided insights into where to focus the detailed circuit analysis efforts. Tasks 8 and 11: Scoping fire modeling and detailed fire modeling. More focus should be placed on detailed fire modeling and less focus on scoping fire modeling. This was the approach taken for the fire PRA. Task 14: Fire risk quantification. Typically, multiple safe shutdown (SSD) components fail during a given fire scenario. Therefore dependent failure analysis is critical to obtaining a meaningful fire risk quantification. Dependent failure analysis for the fire PRA presented several challenges which will be discussed in the full paper.

HTGR accident and risk assessment

International Nuclear Information System (INIS)

Silady, F.A.; Everline, C.J.; Houghton, W.J.

1982-01-01

This paper is a synopsis of the high-temperature gas-cooled reactor probabilistic risk assessments (PRAs) performed by General Atomic Company. Principal topics presented include: HTGR safety assessments, peer interfaces, safety research, process gas explosions, quantitative safety goals, licensing applications of PRA, enhanced safety, investment risk assessments, and PRA design integration

Evaluation of hsp65 Nested PCR-Restriction Analysis (PRA) for Diagnosing Tuberculosis in a High Burden Country

Science.gov (United States)

Macente, Sara; Fujimura Leite, Clarice Queico; Santos, Adolfo Carlos Barreto; Siqueira, Vera Lúcia Dias; Machado, Luzia Neri Cosmo; Marcondes, Nadir Rodrigues; Hirata, Mario Hiroyuki; Hirata, Rosário Dominguez Crespo

2013-01-01

Current study evaluated the hsp65 Nested PCR Restriction Fragment Length Polymorphism Analysis (hsp65 Nested PCR-PRA) to detect and identify Mycobacterium tuberculosis complex directly in clinical samples for a rapid and specific diagnosis of tuberculosis (TB). hsp65 Nested PCR-PRA was applied directly to 218 clinical samples obtained from 127 patients suspected of TB or another mycobacterial infection from July 2009 to July 2010. The hsp65 Nested PCR-PRA showed 100% sensitivity and 95.0 and 93.1% specificity in comparison with culture and microscopy (acid fast bacillus smear), respectively. hsp65 Nested PCR-PRA was shown to be a fast and reliable assay for diagnosing TB, which may contribute towards a fast diagnosis that could help the selection of appropriate chemotherapeutic and early epidemiological management of the cases which are of paramount importance in a high TB burden country. PMID:24260739

Clinical analysis of the changes of plasma PRA, AT-II and Aid levels in patients with acute renal failure

International Nuclear Information System (INIS)

Zhang Qiuyue; Yang Yongqing

2002-01-01

Objective: To investigate the role of changes of plasma PRA, AT-II and Ald levels in the pathogenesis of acute renal failure. Methods: Plasma PRA, AT-II and Ald levels were determined with RIA in 40 normal subjects and 72 cases of acute renal failure. Results: Plasma PRA, AT-II and Ald levels in the patients were markedly increased as compared with those in normal subjects (p < 0.05, p < 0.01, p < 0.001 respectively). There were no linearity and exponential relationship between plasma PRA, AT-II, Ald levels and the 24 h urinary sodium excretion amount (within the range of 89.1 - 365.2 mEq). Conclusion: Acute renal failure could activate the RAAS function

Human reliability analysis in support of a level 1 PRA for Surry during midloop operations

International Nuclear Information System (INIS)

Lin, J.C.; Bley, D.C.; Chu, T.-L.

2004-01-01

The objectives of this Level 1 probabilistic risk assessment (PRA) are to evaluate the important accident sequences initiated during midloop operations and to compare the qualitative and quantitative results with those for accidents initiated during power operations. The primary types of human actions analyzed in this study involve the dynamic operator actions and recovery actions that take place during the accident sequence following an initiating event. Two parts of the human actions were analyzed: failure to diagnose and failure to perform the action. The scope of the Level 1 PRA for Surry during midloop operations includes internal, fire, and flood initiating events. The major categories of dynamic operator actions taken during the accident sequence following an initiating event are: providing makeup to the reactor coolant system (RCS), restoring residual heat removal (RHR) cooling, establishing steam generator reflux cooling, establishing primary feed and spill, establishing gravity feed from refueling water storage tank (RWST), establishing high pressure recirculation, establishing recirculation spray, and cross-connecting RWSTs. All categories are not applicable to all initiating events and all plant operating states (POS). (author)

Validation needs of seismic probabilistic risk assessment (PRA) methods applied to nuclear power plants

International Nuclear Information System (INIS)

Kot, C.A.; Srinivasan, M.G.; Hsieh, B.J.

1985-01-01

An effort to validate seismic PRA methods is in progress. The work concentrates on the validation of plant response and fragility estimates through the use of test data and information from actual earthquake experience. Validation needs have been identified in the areas of soil-structure interaction, structural response and capacity, and equipment fragility. Of particular concern is the adequacy of linear methodology to predict nonlinear behavior. While many questions can be resolved through the judicious use of dynamic test data, other aspects can only be validated by means of input and response measurements during actual earthquakes. A number of past, ongoing, and planned testing programs which can provide useful validation data have been identified, and validation approaches for specific problems are being formulated

An integrated PRA module for fast determination of risk significance and improvement effectiveness

International Nuclear Information System (INIS)

Chao, Chun-Chang; Lin, Jyh-Der

2004-01-01

With the widely use of PRA technology in risk-informed applications, to predict the changes of CDF and LERF becomes a standard process for risk-informed applications. This paper describes an integrated PRA module prepared for risk-informed applications. The module contains a super risk engine, a super fault tree engine, an advanced PRA model and a tool for data base maintenance. The individual element of the module also works well for purpose other than risk-informed applications. The module has been verified and validated through a series of scrupulous benchmark tests with similar software. The results of the benchmark tests showed that the module has remarkable accuracy and speed even for an extremely large-size top-logic fault tree as well as for the case in which large amount of MCSs may be generated. The risk monitor for nuclear power plants in Taiwan is the first application to adopt the module. The results predicted by the risk monitor are now accepted by the regulatory agency. A tool to determine the risk significance according to the inspection findings will be the next application to adopt the module in the near future. This tool classified the risk significance into four different color codes according to the level of increase on CDF. Experience of application showed that the flexibility, the accuracy and speed of the module make it useful in any risk-informed applications when risk indexes must be determined by resolving a PRA model. (author)

Proceedings of the twenty-fourth water reactor safety information meeting. Volume 3: PRA and HRA; Probabilistic seismic hazard assessment and seismic siting criteria

Energy Technology Data Exchange (ETDEWEB)

Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

1997-02-01

This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This volume is divided into the following sections: PRA and HRA and probabilistic seismic hazard assessment and seismic siting criteria. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

Proceedings of the twenty-fourth water reactor safety information meeting. Volume 3: PRA and HRA; Probabilistic seismic hazard assessment and seismic siting criteria

International Nuclear Information System (INIS)

Monteleone, S.

1997-02-01

This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This volume is divided into the following sections: PRA and HRA and probabilistic seismic hazard assessment and seismic siting criteria. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

Clinical evaluation of the new TGDc-01 "PRA" palpebral tonometer: comparison with contact and non-contact tonometry.

Science.gov (United States)

García Resúa, Carlos; Giráldez Fernández, Maria J; Cerviño Expósito, Alejandro; González Pérez, Javier; Yebra-Pimentel, Eva

2005-02-01

The TGDc-01 "PRA" (Ryazan State Instrument, Ryazan, Russia) tonometer is a new portable small-sized tonometer that measures intraocular pressure (IOP) through the eyelid. The purpose of this study is to assess the repeatability of the TGDc-01 IOP measurements by comparing them against those obtained with Goldmann tonometer and with those from Perkins applanation tonometer, Xpert (Reichert, Depew, NY) noncontact tonometer, and Tono-Pen XL (Medtronic Solan, Jacksonville, FL) digital tonometer. Fifty-eight right eyes of 58 young subjects were measured with each of the tonometers. Noncontact tonometry was performed first, followed by Goldmann and Perkins applanation tonometer (in random order), digital Tono-Pen XL, and finally TGDc-01 tonometer (sitting and supine position). Correlation analysis was used to evaluate the relationship between the Goldmann tonometer and the remaining tonometers used in this study. Plotting the difference between the methods against mean was also done to compare the tonometers. The hypothesis of zero bias was examined by a paired t-test. The 95% limits of agreement (LoA) were also calculated. TGDc-01 showed no statistical difference between the IOP measurements obtained in sitting and supine positions. A poor relationship between the TGDc-01 and Goldmann tonometer was found (r = 0.173; p = 0.001). Although the mean differences between Goldmann and Tonopen XL, Xpert, and TGDc-01 IOP measurements were statistically significant, the wider 95% LoA was observed when comparing the Goldmann and TGDc-01 tonometers. Computation of the 95% LoA resulted in a wide bias range when comparing the TGDc-01 with all the tonometers used in this study. The TGDc-01 "PRA" tonometer was not comparable with the other techniques used in the study. The wide dispersion range of the values obtained shows low repeatability of the TGDc-01 for screening purposes. These results could be because of the technique of measurement and/or interindividual variables.

Cognitive environment simulation: An artificial intelligence system for human performance assessment: Cognitive reliability analysis technique: [Technical report, May 1986-June 1987

International Nuclear Information System (INIS)

Woods, D.D.; Roth, E.M.

1987-11-01

This report documents the results of Phase II of a three phase research program to develop and validate improved methods to model the cognitive behavior of nuclear power plant (NPP) personnel. In Phase II a dynamic simulation capability for modeling how people form intentions to act in NPP emergency situations was developed based on techniques from artificial intelligence. This modeling tool, Cognitive Environment Simulation or CES, simulates the cognitive processes that determine situation assessment and intention formation. It can be used to investigate analytically what situations and factors lead to intention failures, what actions follow from intention failures (e.g., errors of omission, errors of commission, common mode errors), the ability to recover from errors or additional machine failures, and the effects of changes in the NPP person-machine system. The Cognitive Reliability Assessment Technique (or CREATE) was also developed in Phase II to specify how CES can be used to enhance the measurement of the human contribution to risk in probabilistic risk assessment (PRA) studies. 34 refs., 7 figs., 1 tab

Survey of seismic fragilities used in PRA studies of nuclear power plants

International Nuclear Information System (INIS)

Park, Y.J.; Hofmayer, C.H.; Chokshi, N.C.

1998-01-01

In recent years, seismic PRA studies have been performed on a large number of nuclear power plants in the USA. This paper presents a summary of a survey on fragility databases and the range of evaluated fragility values of various equipment categories based on past PRAs. The survey includes the use of experience data, the interpretations of available test data, and the quantification of uncertainties. The surveyed fragility databases are limited to data available in the public domain such as NUREG reports, conference proceedings and other publicly available reports. The extent of the availability of data as well as limitations are studied and tabulated for various equipment categories. The survey of the fragility values in past PRA studies includes not only the best estimate values, but also the dominant failure modes and the estimated uncertainty levels for each equipment category. The engineering judgments employed in estimating the uncertainty in the fragility values are also studied. This paper provides a perspective on the seismic fragility evaluation procedures for equipment in order to clearly identify the engineering analysis and judgment used in past seismic PRA studies

Documentation design for probabilistic risk assessment

International Nuclear Information System (INIS)

Parkinson, W.J.; von Herrmann, J.L.

1985-01-01

This paper describes a framework for documentation design of probabilistic risk assessment (PRA) and is based on the EPRI document NP-3470 ''Documentation Design for Probabilistic Risk Assessment''. The goals for PRA documentation are stated. Four audiences are identified which PRA documentation must satisfy, and the documentation consistent with the needs of the various audiences are discussed, i.e., the Summary Report, the Executive Summary, the Main Report, and Appendices. The authors recommend the documentation specifications discussed herein as guides rather than rigid definitions

ATHEANA: open-quotes a technique for human error analysisclose quotes entering the implementation phase

International Nuclear Information System (INIS)

Taylor, J.; O'Hara, J.; Luckas, W.

1997-01-01

Probabilistic Risk Assessment (PRA) has become an increasingly important tool in the nuclear power industry, both for the Nuclear Regulatory Commission (NRC) and the operating utilities. The NRC recently published a final policy statement, SECY-95-126, encouraging the use of PRA in regulatory activities. Human reliability analysis (HRA), while a critical element of PRA, has limitations in the analysis of human actions in PRAs that have long been recognized as a constraint when using PRA. In fact, better integration of HRA into the PRA process has long been a NRC issue. Of particular concern, has been the omission of errors of commission - those errors that are associated with inappropriate interventions by operators with operating systems. To address these concerns, the NRC identified the need to develop an improved HRA method, so that human reliability can be better represented and integrated into PRA modeling and quantification. The purpose of the Brookhaven National Laboratory (BNL) project, entitled 'Improved HRA Method Based on Operating Experience' is to develop a new method for HRA which is supported by the analysis of risk-significant operating experience. This approach will allow a more realistic assessment and representation of the human contribution to plant risk, and thereby increase the utility of PRA. The project's completed, ongoing, and future efforts fall into four phases: (1) Assessment phase (FY 92/93); (2) Analysis and Characterization phase (FY 93/94); (3) Development phase (FY 95/96); and (4) Implementation phase (FY 96/97 ongoing)

Application of determination of PRA, Ang II and IGF-1 levels in the study of typing of essential hypertension

International Nuclear Information System (INIS)

Lu Yongyi; Chen Qun; Yang Yongqing

2010-01-01

Objective: To study the clinical application of determination of plasma renin activity (PRA), Angiotensin II (Ang II ) and insulin-like growth factor-1 (IGF-1) levels in typing of essential hypertension (EH). Methods: Determined the levels of PRA and Aug II in 256 patients with EH and 70 healthy volunteers (as control group) by radioimmunoassay, and measured IGF-1 level by enzyme immunoassay. Research on the typing of EH and the difference between the groups. Results: The PRA and Ang II in control group was (0.432±0.236) μg·L -1 ·h -1 and (31.7±7.4) μg/L respectively. In 256 patients with EH, PRA was increased, normal and decreased in 18.0%, 71.8% and 10.2% respectively, while the level of Ang II was increased, normal and decreased in 12.9%, 76.2% and 10.9% respectively. The IGF-1 levels in 256 patients with EH were increased following the increase of blood pressure. Conclusion: Typing of EH patients with PRA and Ang II as well as the determination of IGF-1 were useful in treating and following up the patients with EH. (authors)

Probabilistic Risk Assessment to Inform Decision Making: Frequently Asked Questions

Science.gov (United States)

General concepts and principles of Probabilistic Risk Assessment (PRA), describe how PRA can improve the bases of Agency decisions, and provide illustrations of how PRA has been used in risk estimation and in describing the uncertainty in decision making.

Licensing topical report: application of probabilistic risk assessment in the selection of design basis accidents

International Nuclear Information System (INIS)

Houghton, W.J.

1980-06-01

A probabilistic risk assessment (PRA) approach is proposed to be used to scrutinize selection of accident sequences. A technique is described in this Licensing Topical Report to identify candidates for Design Basis Accidents (DBAs) utilizing the risk assessment results. As a part of this technique, it is proposed that events with frequencies below a specified limit would not be candidates. The use of the methodology described is supplementary to the traditional, deterministic approach and may result, in some cases, in the selection of multiple failure sequences as DBAs; it may also provide a basis for not considering some traditionally postulated events as being DBAs. A process is then described for selecting a list of DBAs based on the candidates from PRA as supplementary to knowledge and judgments from past licensing practice. These DBAs would be the events considered in Chapter 15 of Safety Analysis Reports of high-temperature gas-cooled reactors

EPRI/NRC-RES fire PRA guide for nuclear power facilities. Volume 1, summary and overview

International Nuclear Information System (INIS)

2004-01-01

This report documents state-of-the-art methods, tools, and data for the conduct of a fire Probabilistic Risk Assessment (PRA) for a commercial nuclear power plant (NPP) application. The methods have been developed under the Fire Risk Re-quantification Study. This study was conducted as a joint activity between EPRI and the U. S. NRC Office of Nuclear Regulatory Research (RES) under the terms of an EPRI/RES Memorandum of Understanding (RS.1) and an accompanying Fire Research Addendum (RS.2). Industry participants supported demonstration analyses and provided peer review of this methodology. The documented methods are intended to support future applications of Fire PRA, including risk-informed regulatory applications. The documented method reflects state-of-the-art fire risk analysis approaches. The primary objective of the Fire Risk Study was to consolidate recent research and development activities into a single state-of-the-art fire PRA analysis methodology. Methodological issues raised in past fire risk analyses, including the Individual Plant Examination of External Events (IPEEE) fire analyses, have been addressed to the extent allowed by the current state-of-the-art and the overall project scope. Methodological debates were resolved through a consensus process between experts representing both EPRI and RES. The consensus process included a provision whereby each major party (EPRI and RES) could maintain differing technical positions if consensus could not be reached. No cases were encountered where this provision was invoked. While the primary objective of the project was to consolidate existing state-of-the-art methods, in many areas, the newly documented methods represent a significant advancement over previously documented methods. In several areas, this project has, in fact, developed new methods and approaches. Such advances typically relate to areas of past methodological debate.

Assessing Probabilistic Risk Assessment Approaches for Insect Biological Control Introductions.

Science.gov (United States)

Kaufman, Leyla V; Wright, Mark G

2017-07-07

The introduction of biological control agents to new environments requires host specificity tests to estimate potential non-target impacts of a prospective agent. Currently, the approach is conservative, and is based on physiological host ranges determined under captive rearing conditions, without consideration for ecological factors that may influence realized host range. We use historical data and current field data from introduced parasitoids that attack an endemic Lepidoptera species in Hawaii to validate a probabilistic risk assessment (PRA) procedure for non-target impacts. We use data on known host range and habitat use in the place of origin of the parasitoids to determine whether contemporary levels of non-target parasitism could have been predicted using PRA. Our results show that reasonable predictions of potential non-target impacts may be made if comprehensive data are available from places of origin of biological control agents, but scant data produce poor predictions. Using apparent mortality data rather than marginal attack rate estimates in PRA resulted in over-estimates of predicted non-target impact. Incorporating ecological data into PRA models improved the predictive power of the risk assessments.

Probabilistic commentary: the rise and fall, and rise again, of risk assessment

International Nuclear Information System (INIS)

Hendrie, J.M.

1985-02-01

Probabilistic risk assessment is mainly concerned with assessing the risks of nuclear power plants. Historically, the field of PRA began with a Senate request for a report on the safety of nuclear reactors in 1972. A quantitative report called WASH-1400 was eventually prepared and published in 1975, and in summary, it stated that nuclear reactors warranted only a low-grade concern in modern society. Criticism of this report and public perception of its results were highly visible subjects in the media, and the criticism led to the fact that PRA fell into disfavor. After Three Mile Island, it was recognized that PRA was a valuable tool for understanding such accidents, and PRA became a bit more popular again by the end of 1979. The usefulness of PRA was also supported by a German study in 1979. PRA played a significant role in the hearings on the Indian Point reactor. The present NRC regards PRA as an important tool in regulatory practice

The development and application of an integrated radiological risk assessment procedure using time-dependent probabilistic risk analysis

International Nuclear Information System (INIS)

Laurens, J.M.; Thompson, B.G.J.; Sumerling, T.J.

1990-01-01

During the past decade, the UKDoE has funded the development of an integrated assessment procedure centred around probabilistic risk analysis (p.r.a.) using Monte Carlo simulation techniques to account for the effects of parameter value uncertainty, including those associated with temporal changes in the environment over a postclosure period of about one million years. The influence of these changes can now be incorporated explicitly into the p.r.a. simulator VANDAL (Variability ANalysis of Disposal ALternatives) briefly described here. Although a full statistically converged time-dependent p.r.a. will not be demonstrated until the current Dry Run 3 trial is complete, illustrative examples are given showing the ability of VANDAL to represent spatially complex groundwater and repository systems evolving under the influence of climatic change. 18 refs., 10 figs., 1 tab

An overview of insights gained and lessons learned from U.S. plant-specific PRA studies

International Nuclear Information System (INIS)

Joksimovich, V.

1985-01-01

Probabilistic Risk Assessment (PRA) has been under development for over twenty years, but it has reached the level of widespread use only in the aftermath of the TMI accident. Over thirty PRAs have now been completed in the U.S. PRAs have been in the mainstream of many licensing decisions because the NRC recognizes that they provide independent and comprehensive plant safety audit. Some difficulties have been experienced leading to interpretive and intercomparison studies. Numerous global and plant-specific insights have been derived. A new application termed risk management is clearly emerging. (orig./HP)

Insights gained through probabilistic risk assessments

International Nuclear Information System (INIS)

Hitchler, M.J.; Burns, N.L.; Liparulo, N.J.; Mink, F.J.

1987-01-01

The insights gained through a comparison of seven probabilistic risk assessments (PRA) studies (Italian PUN, Sizewell B, Ringhals 2, Millstone 3, Zion 1 and 2, Oconee 3, and Seabrook) included insights regarding the adequacy of the PRA technology utilized in the studies and the potential areas for improvement and insights regarding the adequacy of plant designs and how PRA has been utilized to enhance the design and operation of nuclear power plants

The 10,000-year debate

International Nuclear Information System (INIS)

Wilson, J.R.

1996-01-01

Probabilistic Risk Assessment (PRA) has developed into a respected tool within the reactor community. Now, this PRA technique is being applied to a new arena, the distant future of the nuclear waste repository. Problems are already testing the credibility of PRA

The 10,000-year debate

Energy Technology Data Exchange (ETDEWEB)

Wilson, J.R.

1996-08-01

Probabilistic Risk Assessment (PRA) has developed into a respected tool within the reactor community. Now, this PRA technique is being applied to a new arena, the distant future of the nuclear waste repository. Problems are already testing the credibility of PRA.

Assessing Probabilistic Risk Assessment Approaches for Insect Biological Control Introductions

Directory of Open Access Journals (Sweden)

Leyla V. Kaufman

2017-07-01

Full Text Available The introduction of biological control agents to new environments requires host specificity tests to estimate potential non-target impacts of a prospective agent. Currently, the approach is conservative, and is based on physiological host ranges determined under captive rearing conditions, without consideration for ecological factors that may influence realized host range. We use historical data and current field data from introduced parasitoids that attack an endemic Lepidoptera species in Hawaii to validate a probabilistic risk assessment (PRA procedure for non-target impacts. We use data on known host range and habitat use in the place of origin of the parasitoids to determine whether contemporary levels of non-target parasitism could have been predicted using PRA. Our results show that reasonable predictions of potential non-target impacts may be made if comprehensive data are available from places of origin of biological control agents, but scant data produce poor predictions. Using apparent mortality data rather than marginal attack rate estimates in PRA resulted in over-estimates of predicted non-target impact. Incorporating ecological data into PRA models improved the predictive power of the risk assessments.

Comparison of SKIFS 2004:1 and Tillsynshandbok PSA against the ASME PRA Standard and European requirements on PSA

International Nuclear Information System (INIS)

Hellstroem, Per

2005-04-01

Requirements on PSA for risk informed applications are expressed in different international documents. The ASME PRA standard published in spring 2002 is one such document, PSA requirements are also expressed in the European Utility Requirements (EUR) for new reactors. The Swedish PSA requirements are provided in the Swedish regulators (SKI) statutes SKIFS 2004:1. SKI also has a review handbook for PSA activities (SKI report 2003:48). The review handbook is a support during review of the utilities PSA activities and the PSAs themselves. The review handbook expresses SKIs expectations by providing so called important aspects for both the PSA work and the PSAs, A comparison of SKIFS requirements and the important aspects in the Review handbook, on one side, and the requirements on PSA in EUR and ASME on the other side, is presented. The comparison shows a large difference in the level of detail in the different documents, where ASME is most detailed and specific. This is expected since the SKI review handbook not is a 'PSA guide' in the same way as the ASME PRA standard. A direct comparison of the ASME PRA standard requirements with the important aspects in the review handbook cannot answer the question which ASME capacity level that is achieved by a PSA meeting all important aspects. The conclusion is that it is not likely to achieve capacity level 2 and 3, since very few ASME level 3 attributes are explicitly expressed as important aspects, though many are expressed in general terms. The review handbook important aspects that are most similar to the ASME capacity level 1 attributes are initiating events, sequence analysis, and system analysis while less similarity is found for analysis of operator actions data analysis, quantification and containment analysis (level 2). Less similarity is found for capacity level 2 and 3. However, the number of additional ASME attributes on capacity level 2 and 3 are few. There are also important aspects in the review handbook that

A review of NRC staff uses of probabilistic risk assessment

Energy Technology Data Exchange (ETDEWEB)

1994-03-01

The NRC staff uses probabilistic risk assessment (PRA) and risk management as important elements its licensing and regulatory processes. In October 1991, the NRC`s Executive Director for Operations established the PRA Working Group to address concerns identified by the Advisory Committee on Reactor Safeguards with respect to unevenness and inconsistency in the staff`s current uses of PRA. After surveying current staff uses of PRA and identifying needed improvements, the Working Group defined a set of basic principles for staff PRA use and identified three areas for improvements: guidance development, training enhancements, and PRA methods development. For each area of improvement, the Working Group took certain actions and recommended additional work. The Working Group recommended integrating its work with other recent PRA-related activities the staff completed and improving staff interactions with PRA users in the nuclear industry. The Working Group took two key actions by developing general guidance for two uses of PRA within the NRC (that is, screening or prioritizing reactor safety issues and analyzing such issues in detail) and developing guidance on basic terms and methods important to the staff`s uses of PRA.

A review of NRC staff uses of probabilistic risk assessment

International Nuclear Information System (INIS)

1994-03-01

The NRC staff uses probabilistic risk assessment (PRA) and risk management as important elements its licensing and regulatory processes. In October 1991, the NRC's Executive Director for Operations established the PRA Working Group to address concerns identified by the Advisory Committee on Reactor Safeguards with respect to unevenness and inconsistency in the staff's current uses of PRA. After surveying current staff uses of PRA and identifying needed improvements, the Working Group defined a set of basic principles for staff PRA use and identified three areas for improvements: guidance development, training enhancements, and PRA methods development. For each area of improvement, the Working Group took certain actions and recommended additional work. The Working Group recommended integrating its work with other recent PRA-related activities the staff completed and improving staff interactions with PRA users in the nuclear industry. The Working Group took two key actions by developing general guidance for two uses of PRA within the NRC (that is, screening or prioritizing reactor safety issues and analyzing such issues in detail) and developing guidance on basic terms and methods important to the staff's uses of PRA

Use of limited data to construct Bayesian networks for probabilistic risk assessment.

Energy Technology Data Exchange (ETDEWEB)

Groth, Katrina M.; Swiler, Laura Painton

2013-03-01

Probabilistic Risk Assessment (PRA) is a fundamental part of safety/quality assurance for nuclear power and nuclear weapons. Traditional PRA very effectively models complex hardware system risks using binary probabilistic models. However, traditional PRA models are not flexible enough to accommodate non-binary soft-causal factors, such as digital instrumentation&control, passive components, aging, common cause failure, and human errors. Bayesian Networks offer the opportunity to incorporate these risks into the PRA framework. This report describes the results of an early career LDRD project titled %E2%80%9CUse of Limited Data to Construct Bayesian Networks for Probabilistic Risk Assessment%E2%80%9D. The goal of the work was to establish the capability to develop Bayesian Networks from sparse data, and to demonstrate this capability by producing a data-informed Bayesian Network for use in Human Reliability Analysis (HRA) as part of nuclear power plant Probabilistic Risk Assessment (PRA). This report summarizes the research goal and major products of the research.

A comparison of integrated safety analysis and probabilistic risk assessment

International Nuclear Information System (INIS)

Damon, Dennis R.; Mattern, Kevin S.

2013-01-01

The U.S. Nuclear Regulatory Commission conducted a comparison of two standard tools for risk informing the regulatory process, namely, the Probabilistic Risk Assessment (PRA) and the Integrated Safety Analysis (ISA). PRA is a calculation of risk metrics, such as Large Early Release Frequency (LERF), and has been used to assess the safety of all commercial power reactors. ISA is an analysis required for fuel cycle facilities (FCFs) licensed to possess potentially critical quantities of special nuclear material. A PRA is usually more detailed and uses more refined models and data than an ISA, in order to obtain reasonable quantitative estimates of risk. PRA is considered fully quantitative, while most ISAs are typically only partially quantitative. The extension of PRA methodology to augment or supplant ISAs in FCFs has long been considered. However, fuel cycle facilities have a wide variety of possible accident consequences, rather than a few surrogates like LERF or core damage as used for reactors. It has been noted that a fuel cycle PRA could be used to better focus attention on the most risk-significant structures, systems, components, and operator actions. ISA and PRA both identify accident sequences; however, their treatment is quite different. ISA's identify accidents that lead to high or intermediate consequences, as defined in 10 Code of Federal Regulations (CFR) 70, and develop a set of Items Relied on For Safety (IROFS) to assure adherence to performance criteria. PRAs identify potential accident scenarios and estimate their frequency and consequences to obtain risk metrics. It is acceptable for ISAs to provide bounding evaluations of accident consequences and likelihoods in order to establish acceptable safety; but PRA applications usually require a reasonable quantitative estimate, and often obtain metrics of uncertainty. This paper provides the background, features, and methodology associated with the PRA and ISA. The differences between the

Uses of human reliability analysis probabilistic risk assessment results to resolve personnel performance issues that could affect safety

International Nuclear Information System (INIS)

O'Brien, J.N.; Spettell, C.M.

1985-10-01

This report is the first in a series which documents research aimed at improving the usefulness of Probabilistic Risk Assessment (PRA) results in addressing human risk issues. This first report describes the results of an assessment of how well currently available PRA data addresses human risk issues of current concern to NRC. Findings indicate that PRA data could be far more useful in addressing human risk issues with modification of the development process and documentation structure of PRAs. In addition, information from non-PRA sources could be integrated with PRA data to address many other issues. 12 tabs

Expected proton signal sizes in the PRaVDA Range Telescope for proton Computed Tomography

International Nuclear Information System (INIS)

Price, T.; Parker, D.J.; Green, S.; Esposito, M.; Waltham, C.; Allinson, N.M.; Poludniowski, G.; Evans, P.; Taylor, J.; Manolopoulos, S.; Anaxagoras, T.; Nieto-Camero, J.

2015-01-01

Proton radiotherapy has demonstrated benefits in the treatment of certain cancers. Accurate measurements of the proton stopping powers in body tissues are required in order to fully optimise the delivery of such treaments. The PRaVDA Consortium is developing a novel, fully solid state device to measure these stopping powers. The PRaVDA Range Telescope (RT), uses a stack of 24 CMOS Active Pixel Sensors (APS) to measure the residual proton energy after the patient. We present here the ability of the CMOS sensors to detect changes in the signal sizes as the proton traverses the RT, compare the results with theory, and discuss the implications of these results on the reconstruction of proton tracks

Application of probabilistic risk assessment in the operation of Koeberg nuclear power station

International Nuclear Information System (INIS)

Nicholls, D.R.

1991-01-01

Probabilistic risk assessment (PRA) calculates the probability that a set of multiple failures could occur, the frequency with which the safety circuits will be required and the consequences of the failure of the safety systems. In this way the frequency with which major accident situations can be expected to happen, can be derived. The world history of PRA is presented, together with the South African history of PRA. The theory of PRA is explained and the application of PRA studies is described. In the last twenty years, PRA has gone from being a theoretical idea to a practical tool for assisting in plant management. 2 figs., 1 ill

Economic impact assessment in pest risk analysis

NARCIS (Netherlands)

Soliman, T.A.A.; Mourits, M.C.M.; Oude Lansink, A.G.J.M.; Werf, van der W.

2010-01-01

According to international treaties, phytosanitary measures against introduction and spread of invasive plant pests must be justified by a science-based pest risk analysis (PRA). Part of the PRA consists of an assessment of potential economic consequences. This paper evaluates the main available

Probabilistic risk assessment and intelligent decision support systems

International Nuclear Information System (INIS)

Wu, J.S.; Apostolakis, G.E.; Okrent, D.

1989-01-01

The purpose of this paper is to review the progress made in recent years in both the area of PRA as a support to AI applications and the area of AI applications in PRA. The emphasis is on the areas that have made some progress in the past few years, with a brief description of the methods and a discussion of the potential uses and weaknesses. Also included is a brief review of recent developments in the theory of uncertainty in the AI community that may impact uncertainty modeling in PRA. AI techniques could be applied to the related field of PRA in several ways. In this discussion, however, the scope is limited to emergency diagnosis and accident management, because these are the areas that have attracted most of the attention in recent years. The potential use of PRA as a support to these applications is discussed in detail, and this is followed by a survey of recent developments in these areas. (orig./GL)

Current and future applications of PRA in regulatory activities

Energy Technology Data Exchange (ETDEWEB)

Speis, T.P.; Murphy, J.A.; Cunningham, M.A. [Nuclear Regulatory Commission, Washington, DC (United States)] [and others

1995-04-01

Probabilistic Risk Assessments (PRAs) have proven valuable in providing the regulators, the nuclear plant operators, and the reactor designers insights into plant safety, reliability, design and operation. Both the NRC Commissioners and the staff have grown to appreciate the valuable contributions PRAs can have in the regulatory arena, though I will admit the existence of some tendencies for strict adherence to the deterministic approach within the agency and the public at large. Any call for change, particularly one involving a major adjustment in approach to the regulation of nuclear power, will meet with a certain degree of resistance and retrenchment. Change can appear threatening and can cause some to question whether the safety mission is being fulfilled. This skepticism is completely appropriate and is, in fact, essential to a proper transition towards risk and performance-based approaches. Our task in the Office of Nuclear Regulatory Research is to increase the PRA knowledge base within the agency and develop appropriate guidance and methods needed to support the transitioning process.

Probabilistic risk assessment: A look at the role of artificial intelligence

International Nuclear Information System (INIS)

Wang, J.; Modarres, M.; Hunt, R.N.M.

1988-01-01

A review of traditional Probabilistic Risk Assessment (PRA) methods used in the nuclear power industry is presented. The shortcomings of the current PRA methods are pointed out. A method of performing a PRA is proposed and is computerized. The role of artificial intelligence in developing and performing the proposed PRA approach is discussed. The proposed PRA approach is verified by comparing the results to previously performed PRAs. The comparisons have supported the adequacy and completeness of the results of the proposed model. A discussion of how the proposed method can be used as an expert system to verify plant status following loss of plant hardware is also presented. (orig.)

Assessment and presentation of uncertainties in probabilistic risk assessment: how should this be done

International Nuclear Information System (INIS)

Garlick, A.R.; Holloway, N.J.

1987-01-01

Despite continuing improvements in probabilistic risk assessment (PRA) techniques, PRA results, particularly those including degraded core analysis, will have maximum uncertainties of several orders of magnitude. This makes the expression of results, a matter no less important than their estimation. We put forward some ideas on the assessment and expression of highly uncertain quantities, such as probabilities of outcomes of a severe accident. These do not form a consistent set, but rather a number of alternative approaches aimed at stimulating discussion. These include non-probability expressions, such as fuzzy logic or Schafer's support and plausibility which abandon the purely probabilistic expression of risk for a more flexible type of expression, in which other types of measure are possible. The 'risk equivalent plant' concepts represent the opposite approach. Since uncertainty in a risk measure is in itself a form of risk, an attempt is made to define a 'risk equivalent' which is a risk with perfectly defined parameters, regarded (by means of suitable methods of judgement) as 'equally undesirable' with the actual plant. Some guidelines are given on the use of Bayesian methods in data-free or limited data situations. (author)

Use of probabilistic risk assessment in fuel cycle facilities

International Nuclear Information System (INIS)

Gonzalez, Felix; Gonzalez, Michelle; Wagner, Brian

2013-01-01

As expressed in its Policy Statement on the Use of Probabilistic Risk Assessment (PRA) Methods in Nuclear Regulatory Activities, the U.S Nuclear Regulatory Commission has been working for decades to increase the use of PRA technology in its regulatory activities. Since the policy statement was issued in 1995, PRA has become a core component of the nuclear power plant (NPP) licensing and oversight processes. In the last several years, interest has increased in PRA technologies and their possible application to other areas including, but not limited to, spent fuel handling, fuel cycle facilities, reprocessing facilities, and advanced reactors. This paper describes the application of PRA technology currently used in NPPs and its application in other areas such as fuel cycle facilities and advanced reactors. It describes major challenges that are being faced in the application of PRA into new technical areas and possible ways to resolve them. (authors)

Overview of methods for uncertainty analysis and sensitivity analysis in probabilistic risk assessment

International Nuclear Information System (INIS)

Iman, R.L.; Helton, J.C.

1985-01-01

Probabilistic Risk Assessment (PRA) is playing an increasingly important role in the nuclear reactor regulatory process. The assessment of uncertainties associated with PRA results is widely recognized as an important part of the analysis process. One of the major criticisms of the Reactor Safety Study was that its representation of uncertainty was inadequate. The desire for the capability to treat uncertainties with the MELCOR risk code being developed at Sandia National Laboratories is indicative of the current interest in this topic. However, as yet, uncertainty analysis and sensitivity analysis in the context of PRA is a relatively immature field. In this paper, available methods for uncertainty analysis and sensitivity analysis in a PRA are reviewed. This review first treats methods for use with individual components of a PRA and then considers how these methods could be combined in the performance of a complete PRA. In the context of this paper, the goal of uncertainty analysis is to measure the imprecision in PRA outcomes of interest, and the goal of sensitivity analysis is to identify the major contributors to this imprecision. There are a number of areas that must be considered in uncertainty analysis and sensitivity analysis for a PRA: (1) information, (2) systems analysis, (3) thermal-hydraulic phenomena/fission product behavior, (4) health and economic consequences, and (5) display of results. Each of these areas and the synthesis of them into a complete PRA are discussed

Review insights on the probabilistic risk assessment for the Limerick Generating Station

International Nuclear Information System (INIS)

1984-08-01

In recognition of the high population density around the Limerick Generating Station site and the proposed power level, the Philadelphia Electric Company, in response to NRC staff requests, conducted and submitted between March 1981 and November 1983 a probabilistic risk assessment (PRA) on internal event contributors and a severe accident risk assessment on external event contributors to assess risks posed by operation of the plant. The applicant has developed perspectives using PRA models on the safety profile of the Limerick plant and has altered the plant design to reduce accident vulnerabilities identified in these PRAs. The staff's review of the Limerick PRA has particularly emphasized the dominant accident sequences and the resulting insights into demonstration of compliance with regulatory requirments, unique design features and major plant vulnerabilities to assess the need for any additional measures to further improve the safety of the LGS. The staff's review insights and PRA safety review conclusions are presented in this report

Integrated Level 3 risk assessment for the LaSalle Unit 2 nuclear power plant

International Nuclear Information System (INIS)

Payne, A.C. Jr.; Brown, T.D.; Miller, L.A.

1991-01-01

An integrated Level 3 probabilistic risk assessment (PRA) was performed on the LaSalle County Station nuclear power plant using state-of-the-art PRA analysis techniques. The objective of this study was to provide an estimate of the risk to the offsite population during full power operation of the plant and to include a characterization of the uncertainties in the calculated risk values. Uncertainties were included in the accident frequency analysis, accident progression analysis, and the source term analysis. Only weather uncertainties were included in the consequence analysis. In this paper selected results from the accident frequency, accident progression, source term, consequence, and integrated risk analyses are discussed and the methods used to perform a fully integrated Level 3 PRA are examined. LaSalle County Station is a two-unit nuclear power plant located 55 miles southwest of Chicago, Illinois. Each unit utilizes a Mark 2 containment to house a General Electric 3323 MWt BWR-5 reactor. This PRA, which was performed on Unit 2, included internal as well as external events. External events that were propagated through the risk analysis included earthquakes, fires, and floods. The internal event accident scenarios included transients, transient-induced LOCAs (inadvertently stuck open relief valves), anticipated transients without scram, and loss of coolant accidents

Revision of the AESJ Standard for Seismic Probabilistic Risk Assessment (PRA). Updating requirements based on the lessons learned from the Fukushima Dai-ichi NPP Accidents (3). Fragility evaluation and outline of the updated points

International Nuclear Information System (INIS)

Yamaguchi, Akira; Nakamura, Susumu; Mihara, Yoshinori

2014-01-01

Lessons learned from Great East Japan earthquake and other new findings had been accumulated on the fragility evaluation of buildings and components. And also new analysis and evaluation method had been proposed with the advancement of recent analysis and evaluation technology. These were reflected in revision of the AESJ Standard for Seismic Probabilistic Risk Assessment (PRA). Scope of the fragility evaluation were extended to all equipment on the site, severe accident management equipment including portable equipment and earthquake concomitant incident (such as tsunami) countermeasure equipment. This article described outlines of updating points of the fragility evaluation of the AESJ Standard for Seismic PRA; (1) requirements for seismic induced other risk evaluations such as fire, inundation and tsunami, (2) simulation technology based on recent findings such as three dimensional responses of buildings / structures and its effect on equipment, (3) requirements of the fragility evaluation for various failure mode of several equipment such as severe accident management equipment, fine failure mode of buildings / structures, failures of equipment related with earthquake concomitant incidents (embankment and seawall) and spent fuel pool, and (4) requirements for the fragility evaluation of aftershocks and soil deformation due to fault displacement. (T. Tanaka)

Development of fire PRA methodologies for the analysis of typical Italian NPP designs

International Nuclear Information System (INIS)

Silvestri, E.; Dore, B.; Ferro, G.; Apostolakis, G.

1987-01-01

To compute fire induced Core Melt probability, the results of hazard and propagation analyses were combined with the Core Melt frequency computed for the initiating event and the support state as determined by the fire considered. From the PRA for internal event, the average value of this frequency was found 2.5x10 -3 event/year. Using the average fire frequency the resulting fire induced Core Melt frequency is 1.4x10 -8 event/year. Although high separation of safety systems is required in Italian PWR plants, the frequency of fire induced Core Melt can reach values not negligible with respect to Italian safety standards. For this reason, fire PRA studies for the entire plant are considered necessary and should be performed with appropriate modifications of the methods used for the American plants in order to be able to estimate lower fire induced Core Melt frequencies. (orig./HP)

An Approach to On-line Risk Assessment in NPP

International Nuclear Information System (INIS)

Simic, Z.; Mikulicic, V.; O'Brien, J.

1996-01-01

Probabilistic Risk Assessment (PRA) can provide safety status information for a plant during different configurations; additional effort is needed to do this in real time for on-line operation. This paper describes an approach to use PRA to achieve these goals. A Risk Assessment On-Line (RAOL) application was developed to monitor maintenance (on-line and planned) activities. RAOL is based on the results from a full-scope PRA, engineering/operational judgment and incorporates a user friendly program interface approach. Results from RAOL can be used by planners or operations to effectively manage the level of risk by controlling the actual plant configuration. (author)

Applications of the EBR-II Probabilistic Risk Assessment

International Nuclear Information System (INIS)

Roglans, J.: Ragland, W.A.; Hill, D.J.

1993-01-01

A Probabilistic Risk Assessment (PRA) of the Experimental Breeder Reactor 11 (EBR-11), a Department of Energy (DOE) Category A research reactor, has recently been completed at Argonne National Laboratory (ANL), and has been performed with close collaboration between PRA analysts and engineering and operations staff. A product of this Involvement of plant personnel has been a excellent acceptance of the PRA as a tool, which has already resulted In a variety of applications of the EBR-11 PRA. The EBR-11 has been used in support of plant hardware and procedure modifications and In new system design work. A new application in support of the refueling safety analysis will be completed in the near future

Technical requirements for the ASME PRA standard for nuclear power plant applications

International Nuclear Information System (INIS)

Fleming, Karl N.; Bernsen, Sidney A.; Simard, Ronald L.

2000-01-01

In 1998 the American Society of Mechanical Engineers (ASME) formed the Committee on Nuclear Risk Management (CNRM) and a Project Team to develop a standard on PRAs for use in risk informed applications. This ASME standard is being developed to help provide an adequate level of quality in PRAs that are being used to support ASME initiatives to risk informed in-service inspection (ISI) and in-service testing (IST) of nuclear power plant components. A related need supported by the industry and the U.S. Nuclear Regulatory Commission is to reduce the level of effort that is being expended in pilot applications of risk informed initiatives to address questions about the sufficiency of quality in the supporting PRA models. The purpose of this paper is to discuss the authors' views on some of the technical issues that were encountered in the effort to develop the ASME PRA standard. Draft 12 of this standard has been issued for comment, and is currently being finalized with the aim of releasing the standard in early 2001. (author)

Examination of Conservatism in Early/Latent Fatality Estimation in Level 3 PRA

International Nuclear Information System (INIS)

Kim, Sung-yeop; Lee, Haneol; Yim, Man-Sung

2014-01-01

Due to the computational model driven-nature of the work, there exist various sources of uncertainty in level 3 PRA. They are related with source release, environmental transport and deposition, human behavior involved in dosimetry, health effect and risk assessment. For instance, a total of 376 parameters have been considered in Probabilistic Accident Consequence Uncertainty Assessment Using COSYMA and the details on the number of parameters in each analysis are listed in Table 1. In 2012, the report of NPP accident consequence simulation was distributed by the Korean Federation for Environmental Movement (KFEM). They insisted that Kori Nuclear Power Plant (NPP) accident would lead to 48,000 early fatalities and 850,000 cancer fatalities in Busan and Hanbit NPP accident would lead to 550,000 cancer fatalities in Seoul. This report exemplifies the misuse of collective dose, that is effective dose multiplied by population and time. Even though very low effective dose is considered, collective dose could give over-conservative estimate when high population and long time period is multiplied. International Commission on Radiological Protection (ICRP) forewarned about the misuse of collective dose, in their ICRP Publication 103, such as applying it to simplified calculation of fatality and risk. As part of investigation of conservatism in early and latent fatality estimation, the existing methods of early and latent fatality calculation was reviewed and the results from the use of the existing methodology were examined in this study. The method of early and latent fatality estimation in level 3 PRA was investigated and the conservatism in the result was examined in this study. For the purpose of estimating both early and latent fatality, appropriate dose distributions among the affected population are found to be important. This study showed that large conservatism may be involved in the estimated fatality if the distribution of population dose as a function of

Examination of Conservatism in Early/Latent Fatality Estimation in Level 3 PRA

Energy Technology Data Exchange (ETDEWEB)

Kim, Sung-yeop; Lee, Haneol; Yim, Man-Sung [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2014-10-15

Due to the computational model driven-nature of the work, there exist various sources of uncertainty in level 3 PRA. They are related with source release, environmental transport and deposition, human behavior involved in dosimetry, health effect and risk assessment. For instance, a total of 376 parameters have been considered in Probabilistic Accident Consequence Uncertainty Assessment Using COSYMA and the details on the number of parameters in each analysis are listed in Table 1. In 2012, the report of NPP accident consequence simulation was distributed by the Korean Federation for Environmental Movement (KFEM). They insisted that Kori Nuclear Power Plant (NPP) accident would lead to 48,000 early fatalities and 850,000 cancer fatalities in Busan and Hanbit NPP accident would lead to 550,000 cancer fatalities in Seoul. This report exemplifies the misuse of collective dose, that is effective dose multiplied by population and time. Even though very low effective dose is considered, collective dose could give over-conservative estimate when high population and long time period is multiplied. International Commission on Radiological Protection (ICRP) forewarned about the misuse of collective dose, in their ICRP Publication 103, such as applying it to simplified calculation of fatality and risk. As part of investigation of conservatism in early and latent fatality estimation, the existing methods of early and latent fatality calculation was reviewed and the results from the use of the existing methodology were examined in this study. The method of early and latent fatality estimation in level 3 PRA was investigated and the conservatism in the result was examined in this study. For the purpose of estimating both early and latent fatality, appropriate dose distributions among the affected population are found to be important. This study showed that large conservatism may be involved in the estimated fatality if the distribution of population dose as a function of

Relay chatter and operator response after a large earthquake: An improved PRA methodology with case studies

International Nuclear Information System (INIS)

Budnitz, R.J.; Lambert, H.E.; Hill, E.E.

1987-08-01

The purpose of this project has been to develop and demonstrate improvements in the PRA methodology used for analyzing earthquake-induced accidents at nuclear power reactors. Specifically, the project addresses methodological weaknesses in the PRA systems analysis used for studying post-earthquake relay chatter and for quantifying human response under high stress. An improved PRA methodology for relay-chatter analysis is developed, and its use is demonstrated through analysis of the Zion-1 and LaSalle-2 reactors as case studies. This demonstration analysis is intended to show that the methodology can be applied in actual cases, and the numerical values of core-damage frequency are not realistic. The analysis relies on SSMRP-based methodologies and data bases. For both Zion-1 and LaSalle-2, assuming that loss of offsite power (LOSP) occurs after a large earthquake and that there are no operator recovery actions, the analysis finds very many combinations (Boolean minimal cut sets) involving chatter of three or four relays and/or pressure switch contacts. The analysis finds that the number of min-cut-set combinations is so large that there is a very high likelihood (of the order of unity) that at least one combination will occur after earthquake-caused LOSP. This conclusion depends in detail on the fragility curves and response assumptions used for chatter. Core-damage frequencies are calculated, but they are probably pessimistic because assuming zero credit for operator recovery is pessimistic. The project has also developed an improved PRA methodology for quantifying operator error under high-stress conditions such as after a large earthquake. Single-operator and multiple-operator error rates are developed, and a case study involving an 8-step procedure (establishing feed-and-bleed in a PWR after an earthquake-initiated accident) is used to demonstrate the methodology

Role of seismic PRA in seismic safety decisions of nuclear power plants

International Nuclear Information System (INIS)

Ravindra, M.K.; Kennedy, R.P.; Sues, R.H.

1985-01-01

This paper highlights the important roles that seismic probabilistic risk assessments (PRAs) can play in the seismic safety decisions of nuclear power plants. If a seismic PRA has been performed for a plant, its results can be utilized to evaluate the seismic capability beyond the safe shutdown event (SSE). Seismic fragilities of key structures and equipment, fragilities of dominant plant damage states and the frequencies of occurrence of these plant damage states are reviewed to establish the seismic safety of the plant beyond the SSE level. Guidelines for seismic margin reviews and upgrading may be developed by first identifying the generic classes of structures and equipment that have been shown to be dominant risk contributors in the completed seismic PRAs, studying the underlying causes for their contribution and examining why certain other items (e.g., piping) have not proved to be high-risk-contributors

Use of Probabilistic Risk Assessment in Shuttle Decision Making Process

Science.gov (United States)

Boyer, Roger L.; Hamlin, Teri, L.

2011-01-01

This slide presentation reviews the use of Probabilistic Risk Assessment (PRA) to assist in the decision making for the shuttle design and operation. Probabilistic Risk Assessment (PRA) is a comprehensive, structured, and disciplined approach to identifying and analyzing risk in complex systems and/or processes that seeks answers to three basic questions: (i.e., what can go wrong? what is the likelihood of these occurring? and what are the consequences that could result if these occur?) The purpose of the Shuttle PRA (SPRA) is to provide a useful risk management tool for the Space Shuttle Program (SSP) to identify strengths and possible weaknesses in the Shuttle design and operation. SPRA was initially developed to support upgrade decisions, but has evolved into a tool that supports Flight Readiness Reviews (FRR) and near real-time flight decisions. Examples of the use of PRA for the shuttle are reviewed.

Comments of the PRA Senior Review Panel on the meeting held December 1--3, 1987

International Nuclear Information System (INIS)

Sharp, D.A.

1988-01-01

This memorandum records the minutes of the PRA Senior Review Panel meeting held at Savannah River Laboratory (SRL) on December 1--3, 1987, and the report on that meeting written subsequently by the panel members. The minutes are contained as Attachment 2 of this memorandum, and the report as Attachment 1. The Panel indicated two principal concerns in their report: (1) that insufficient emphasis is being placed on the reliability data development program, and (2) that excessive detail is being built into the fault trees. These concerns have been addressed in a subsequent meeting with the Panel, held March 2--4, 1988. In addition, the members have been provided with a program document (Reference 1) indicating the extent, the timing, and the limitations of the data analysis effort for the PRA

Risk-Informed Safety Assurance and Probabilistic Assessment of Mission-Critical Software-Intensive Systems

Science.gov (United States)

Guarro, Sergio B.

2010-01-01

This report validates and documents the detailed features and practical application of the framework for software intensive digital systems risk assessment and risk-informed safety assurance presented in the NASA PRA Procedures Guide for Managers and Practitioner. This framework, called herein the "Context-based Software Risk Model" (CSRM), enables the assessment of the contribution of software and software-intensive digital systems to overall system risk, in a manner which is entirely compatible and integrated with the format of a "standard" Probabilistic Risk Assessment (PRA), as currently documented and applied for NASA missions and applications. The CSRM also provides a risk-informed path and criteria for conducting organized and systematic digital system and software testing so that, within this risk-informed paradigm, the achievement of a quantitatively defined level of safety and mission success assurance may be targeted and demonstrated. The framework is based on the concept of context-dependent software risk scenarios and on the modeling of such scenarios via the use of traditional PRA techniques - i.e., event trees and fault trees - in combination with more advanced modeling devices such as the Dynamic Flowgraph Methodology (DFM) or other dynamic logic-modeling representations. The scenarios can be synthesized and quantified in a conditional logic and probabilistic formulation. The application of the CSRM method documented in this report refers to the MiniAERCam system designed and developed by the NASA Johnson Space Center.

Applicability of PRISM PRA Methodology to the Level II Probabilistic Safety Analysis of KALIMER-600 (I) (Core Damage Event Tree Analysis Part)

International Nuclear Information System (INIS)

Park, S. Y.; Kim, T. W.; Ha, K. S.; Lee, B. Y.

2009-03-01

The Korea Atomic Energy Research Institute (KAERI) has been developing liquid metal reactor (LMR) design technologies under a National Nuclear R and D Program. Nevertheless, there is no experience of the PSA domestically for a fast reactor with the metal fuel. Therefore, the objective of this study is to establish the methodologies of risk assessment for the reference design of KALIMER-600 reactor. An applicability of the PSA of the PRISM plant to the KALIMER-600 has been studied. The study is confined to a core damage event tree analysis which is a part of a level 2 PSA. Assuming that the accident types, which can be developed from level 1 PSA, are same as the PRISM PRA, core damage categories are defined and core damage event trees are developed for the KALIMER-600 reactor. Fission product release fractions of the core damage categories and branch probabilities of the core damage event trees are referred from the PRISM PRA temporarily. Plant specific data will be used during the detail analysis

Reliability and Probabilistic Risk Assessment - How They Play Together

Science.gov (United States)

Safie, Fayssal M.; Stutts, Richard G.; Zhaofeng, Huang

2015-01-01

PRA methodology is one of the probabilistic analysis methods that NASA brought from the nuclear industry to assess the risk of LOM, LOV and LOC for launch vehicles. PRA is a system scenario based risk assessment that uses a combination of fault trees, event trees, event sequence diagrams, and probability and statistical data to analyze the risk of a system, a process, or an activity. It is a process designed to answer three basic questions: What can go wrong? How likely is it? What is the severity of the degradation? Since 1986, NASA, along with industry partners, has conducted a number of PRA studies to predict the overall launch vehicles risks. Planning Research Corporation conducted the first of these studies in 1988. In 1995, Science Applications International Corporation (SAIC) conducted a comprehensive PRA study. In July 1996, NASA conducted a two-year study (October 1996 - September 1998) to develop a model that provided the overall Space Shuttle risk and estimates of risk changes due to proposed Space Shuttle upgrades. After the Columbia accident, NASA conducted a PRA on the Shuttle External Tank (ET) foam. This study was the most focused and extensive risk assessment that NASA has conducted in recent years. It used a dynamic, physics-based, integrated system analysis approach to understand the integrated system risk due to ET foam loss in flight. Most recently, a PRA for Ares I launch vehicle has been performed in support of the Constellation program. Reliability, on the other hand, addresses the loss of functions. In a broader sense, reliability engineering is a discipline that involves the application of engineering principles to the design and processing of products, both hardware and software, for meeting product reliability requirements or goals. It is a very broad design-support discipline. It has important interfaces with many other engineering disciplines. Reliability as a figure of merit (i.e. the metric) is the probability that an item will

Sequence variations and protein expression levels of the two immune evasion proteins Gpm1 and Pra1 influence virulence of clinical Candida albicans isolates.

Science.gov (United States)

Luo, Shanshan; Hipler, Uta-Christina; Münzberg, Christin; Skerka, Christine; Zipfel, Peter F

2015-01-01

Candida albicans, the important human fungal pathogen uses multiple evasion strategies to control, modulate and inhibit host complement and innate immune attack. Clinical C. albicans strains vary in pathogenicity and in serum resistance, in this work we analyzed sequence polymorphisms and variations in the expression levels of two central fungal complement evasion proteins, Gpm1 (phosphoglycerate mutase 1) and Pra1 (pH-regulated antigen 1) in thirteen clinical C. albicans isolates. Four nucleotide (nt) exchanges, all representing synonymous exchanges, were identified within the 747-nt long GPM1 gene. For the 900-nt long PRA1 gene, sixteen nucleotide exchanges were identified, which represented synonymous, as well as non-synonymous exchanges. All thirteen clinical isolates had a homozygous exchange (A to G) at position 73 of the PRA1 gene. Surface levels of Gpm1 varied by 8.2, and Pra1 levels by 3.3 fold in thirteen tested isolates and these differences influenced fungal immune fitness. The high Gpm1/Pra1 expressing candida strains bound the three human immune regulators more efficiently, than the low expression strains. The difference was 44% for Factor H binding, 51% for C4BP binding and 23% for plasminogen binding. This higher Gpm1/Pra1 expressing strains result in enhanced survival upon challenge with complement active, Factor H depleted human serum (difference 40%). In addition adhesion to and infection of human endothelial cells was increased (difference 60%), and C3b surface deposition was less effective (difference 27%). Thus, variable expression levels of central immune evasion protein influences immune fitness of the human fungal pathogen C. albicans and thus contribute to fungal virulence.

Percutaneous transluminal angioplasty in the region of the renal artery: indications, technique and results

Energy Technology Data Exchange (ETDEWEB)

Olbert, F; Ogris, E; Muzika, N; Schlegl, A; Vacariu, O; Diez, W

1985-01-01

The paper reports the results obtained by percutaneous transluminal angioplasty (PTA) in the treatment of renal artery stenosis in 25 patients with secondary hypertension and discusses indications and the technique used. The follow-up period ranged from 0.5 to 12 months. All interventions were performed with the Olbert catheter system. Following PTA a significant decrease in systolic, diastolic and mean arterial pressure was observed in nearly all patients. The response of the plasma renin activity (PRA) to orthostatic stimulation and pharmacological stimulation by furosemide was determined in a subset of 15 patients before and after PTA. A significant decrease in both basal and stimulated PRA values was observed subsequent to PTA. The decrease in PRA values after successful PTA did not, however, correlate with the extent of the blood pressure changes. (Author).

Percutaneous transluminal angioplasty in the region of the renal artery: indications, technique and results

International Nuclear Information System (INIS)

Olbert, F.; Ogris, E.; Muzika, N.; Schlegl, A.; Vacariu, O.; Diez, W.

1985-01-01

The paper reports the results obtained by percutaneous transluminal angioplasty (PTA) in the treatment of renal artery stenosis in 25 patients with secondary hypertension and discusses indications and the technique used. The follow-up period ranged from 0.5 to 12 months. All interventions were performed with the Olbert catheter system. Following PTA a significant decrease in systolic, diastolic and mean arterial pressure was observed in nearly all patients. The response of the plasma renin activity (PRA) to orthostatic stimulation and pharmacological stimulation by furosemide was determined in a subset of 15 patients before and after PTA. A significant decrease in both basal and stimulated PRA values was observed subsequent to PTA. The decrease in PRA values after successful PTA did not, however, correlate with the extent of the blood pressure changes. (Author)

Can we trust PRA?

Energy Technology Data Exchange (ETDEWEB)

Epstein, S. [ABS Consulting, Koraku Mori, Building, 1-4-14 Koraku Chome, Bunkyo-ku, Tokyo 112-0004 (Japan)]. E-mail: sepstein@absconsulting.com; Rauzy, A. [Institut de Mathematique de Luminy (IML/CNRS), 163 Avenue de Luminy, Case 907, Marseille, Cedex 913288 (France)]. E-mail: arauzy@iml.univ-mrs.fr

2005-06-01

The Fault-Tree/Event-Tree method is widely used in industry as the underlying formalism of probabilistic risk assessment. Almost all of the tools available to assess Event-Tree models implement the 'classical' assessment technique based on minimal cutsets and the rare event approximation. Binary decision diagrams (BDDs) are an alternative approach, but they were up to now limited to medium size models because of the exponential blow up of the memory requirements. We have designed a set of heuristics, which make it possible to quantify, by means of BDD, all of the sequences of a large Event-Tree model coming from the nuclear industry. For the first time, it was possible to compare results of the classical approach with those of the BDD approach, i.e. with exact results. This article reports this comparison and shows that the minimal cutsets technique gives overestimated results in a significant proportion of cases and underestimated results in some cases as well. Hence, the (indeed provocative) question in the title of this article.

Pulsa o coração da cidade: errâncias, afectos e potências no dia e na noite da Praça do Ferreira

Directory of Open Access Journals (Sweden)

Alice Dote

2017-12-01

Full Text Available O presente artigo aborda a potência dos usos, contra-usos e modos de habitar dos artistas de rua da Praça do Ferreira, na cidade de Fortaleza, Ceará. O trabalho apoia-se nos percursos e nas errâncias urbanas da vivência na e da Praça do Ferreira em diferentes temporalidades (diurna e noturna, especialmente no contexto de apresentações noturnas do Grupo As 10 Graças de Palhaçaria aos moradores da Praça. Através desses que têm a rua como casa, agem pelas brechas e proliferam-se pelas margens, proponho-me a perceber a potência da arte de rua, do encontro e da experiência de alteridade na Praça do Ferreira. Finalizo o texto apontando que esse local, assim ocupado, se impregna de significados outros e revela-se como um território de criação, de inventividade, de existência e resistência, portanto, de potência de vida que é, em si, potência política. Palavras-chave: Praça do Ferreira; Fortaleza; cidade; arte urbana; artista de rua

Bruce NGS B risk assessment (BBRA) peer review process

International Nuclear Information System (INIS)

Kaasalainen, S.; Crocker, W.P.; Webb, W.A.

2001-01-01

Risk-informed decision making is considered an effective approach to managing the risk of nuclear power plant operation in a competitive market. Hence, increased reliance on the station probabilistic risk assessments (PRAs) to provide risk perspective inputs is inevitable. With increased reliance on the PRAs it is imperative that PRAs have the characteristics necessary to provide the required information. Recognizing the increased requirements on nuclear power plant PRAs the nuclear industry in the United States has expended significant effort over the past few years defining the required characteristics of a PRA for various applications. More recently several owners groups have drafted guidelines for PRA certification and several U.S. utilities have had their PRAs certified. During the year 2000 Ontario Power Generation, Nuclear (OPG,N) subjected the PRA of one of its stations to the U.S. style certification process. The PRA selected for this process was the Bruce B Risk Assessment (BBRA). BBRA was chosen for this process since it is the first OPG, N PRA to be used for risk-informed applications. However, the strengths of the BBRA identified from the certification process and the lessons learned are also largely applicable to the other OPG, N plant PRAs due to the use of similar methods and tools

Effect of antigravity suit inflation on cardiovascular, PRA, and PVP responses in humans.

Science.gov (United States)

Kravik, S E; Keil, L C; Geelen, G; Wade, C E; Barnes, P R; Spaul, W A; Elder, C A; Greenleaf, J E

1986-08-01

Blood pressure, pulse rate (PR), serum osmolality and electrolytes, as well as plasma vasopressin (PVP) and plasma renin activity (PRA), were measured in five men and two women [mean age 38.6 +/- 3.9 (SE) yr] before, during, and after inflation of an antigravity suit that covered the legs and abdomen. After 24 h of fluid deprivation the subjects stood quietly for 3 h: the 1st h without inflation, the 2nd with inflation to 60 Torr, and the 3rd without inflation. A similar control noninflation experiment was conducted 10 mo after the inflation experiment using five of the seven subjects except that the suit was not inflated during the 3-h period. Mean arterial pressure increased by 14 +/- 4 (SE) Torr (P less than 0.05) with inflation and decreased by 15 +/- 5 Torr (P less than 0.05) after deflation. Pulse pressure (PP) increased by 7 +/- 2 Torr (P less than 0.05) with inflation and PR decreased by 11 +/- 5 beats/min (P less than 0.05); PP and PR returned to preinflation levels after deflation. Plasma volume decreased by 6.1 +/- 1.5% and 5.3 +/- 1.6% (P less than 0.05) during hours 1 and 3, respectively, and returned to base line during inflation. Inflation decreased PVP from 6.8 +/- 1.1 to 5.6 +/- 1.4 pg/ml (P less than 0.05) and abolished the significant rise in PRA during hour 1. Both PVP and PRA increased significantly after deflation: delta = 18.0 +/- 5.1 pg/ml and 4.34 +/- 1.71 ng angiotensin I X ml-1 X h-1, respectively. Serum osmolality and Na+ and K+ concentrations were unchanged during the 3 h of standing.(ABSTRACT TRUNCATED AT 250 WORDS)

Development of a methodology for post closure radiological risk analysis of underground waste repositories. Illustrative assessment of the Harwell site

International Nuclear Information System (INIS)

Gralewski, Z.A.; Kane, P.; Nicholls, D.B.

1987-06-01

A probabilistic risk analysis (pra) is demonstrated for a number of ground water mediated release scenarios at the Harwell Site for a hypothetical repository at a depth of about 150 metres. This is the second stage of development of an overall risk assessment methodology. A procedure for carrying out multi-scenario assessment using available probabilistic risk assessment (pra) models is presented and a general methodology for combining risk contributions is outlined. Appropriate levels of model complexity in pra are discussed. Modelling requirements for the treatment of multiple simultaneous pathways and of site evolution are outlined. Further developments of pra systems are required to increase the realism of both the models and their mode of application, and hence to improve estimates of risk. (author)

Review process and quality assurance in the EBR-II probabilistic risk assessment

International Nuclear Information System (INIS)

Roglans, J.; Hill, D.J.; Ragland, W.A.

1992-01-01

A Probabilistic Risk Assessment (PRA) of the Experimental Breeder Reactor II (EBR-II), a Department of Energy (DOE) Category A reactor, has recently been completed at Argonne National Laboratory (ANL). Within the scope of the ANL QA Programs, a QA Plan specifically for the EBR-II PRA was developed. The QA Plan covered all aspects of the PRA development, with emphasis on the procedures for document and software control, and the internal and external review process. The effort spent in the quality assurance tasks for the EBR-II PRA has reciprocated by providing acceptance of the work and confidence in the quality of the results

Implications of probabilistic risk assessment

International Nuclear Information System (INIS)

Cullingford, M.C.; Shah, S.M.; Gittus, J.H.

1987-01-01

Probabilistic risk assessment (PRA) is an analytical process that quantifies the likelihoods, consequences and associated uncertainties of the potential outcomes of postulated events. Starting with planned or normal operation, probabilistic risk assessment covers a wide range of potential accidents and considers the whole plant and the interactions of systems and human actions. Probabilistic risk assessment can be applied in safety decisions in design, licensing and operation of industrial facilities, particularly nuclear power plants. The proceedings include a review of PRA procedures, methods and technical issues in treating uncertainties, operating and licensing issues and future trends. Risk assessment for specific reactor types or components and specific risks (eg aircraft crashing onto a reactor) are used to illustrate the points raised. All 52 articles are indexed separately. (U.K.)

Bayesian parameter estimation in probabilistic risk assessment

International Nuclear Information System (INIS)

Siu, Nathan O.; Kelly, Dana L.

1998-01-01

Bayesian statistical methods are widely used in probabilistic risk assessment (PRA) because of their ability to provide useful estimates of model parameters when data are sparse and because the subjective probability framework, from which these methods are derived, is a natural framework to address the decision problems motivating PRA. This paper presents a tutorial on Bayesian parameter estimation especially relevant to PRA. It summarizes the philosophy behind these methods, approaches for constructing likelihood functions and prior distributions, some simple but realistic examples, and a variety of cautions and lessons regarding practical applications. References are also provided for more in-depth coverage of various topics

Review of KSNP LPSD PSA model based of ANS LPSD PRA standard, rev.0

International Nuclear Information System (INIS)

Jang, S. C.; Park, J. H.; Kim, T. W.; Lim, H. G.; Yang, J. E.; Ha, J. J.

2004-02-01

Recently, under the de-regulation environment, nuclear industry has attempted various approaches to improve the economics of Nuclear Power Plants (NPP). One of these efforts is the Risk Informed/Performance-based Operation (RIPBO). This approach uses the risk and performance information to manage the resources effectively and efficiently that are used in the operation of NPP. In RIPBO, PSA quality is one of the most important things. The nuclear industry and regulatory body of U.S.A have developed a measure to evaluate the quality of PSA. NEI (Nuclear Energy Institute) has developed a guidance called 'NEI PRA Peer Review Guidance,' and NRC (Nuclear Regulatory Committee) and ASME have developed the 'PRA Standard.' In Korea, several projects are on going now, such as the extension of AOT/STI of RPS/ESFAS, Risk-informed In-service Inspection (RI-ISI). However, in Korea, there have been no attempts to evaluate the quality of PSA model itself. Therefore, we cannot be sure about the quality of PSA whether or not the present PSA model can be used for the risk-informed applications such as mentioned above. We can say that the evaluation of PSA model quality is the basis for the RIPBO. In this report, we have evaluated the quality of PSA model at Low power and Shutdown operation model for Yongkwang 5 and 6 units based on the ANS LPSD PRA Standard. We, also, have derived what items are to be improved to upgrade the quality of LPSD PSA model and how it can be improved. This report can be used as the base of RIPBO work in Korea

Probabilistic risk assessment of HTGRs

International Nuclear Information System (INIS)

Fleming, K.N.; Houghton, W.J.; Hannaman, G.W.; Joksimovic, V.

1980-08-01

Probabilistic Risk Assessment methods have been applied to gas-cooled reactors for more than a decade and to HTGRs for more than six years in the programs sponsored by the US Department of Energy. Significant advancements to the development of PRA methodology in these programs are summarized as are the specific applications of the methods to HTGRs. Emphasis here is on PRA as a tool for evaluating HTGR design options. Current work and future directions are also discussed

Development of a methodology for post closure radiological risk analysis of underground waste repositories. Illustrative assessment of the Harwell site. V.1

International Nuclear Information System (INIS)

Gralewski, Z.A.; Kane, P.; Nicholls, D.B.

1987-06-01

A probabilistic risk analysis (pra) is demonstrated for a number of ground water mediated release scenarios at the Harwell Site for a hypothetical repository at a depth of about 150 metres. This is the second stage of development of an overall risk assessment methodology. A procedure for carrying out multi-scenario assessment using available probabilistic risk assessment (pra) models is presented and a general methodology for combining risk contributions is outlined. Appropriate levels of model complexity in pra are discussed. Modelling requirements for the treatment of multiple simultaneous pathways and of site evolution are outlined. Further developments of pra systems are required to increase the realism of both the models and their mode of application, and hence to improve estimates of risk. (author)

Probabilistic risk assessment of HTGRs

International Nuclear Information System (INIS)

Fleming, K.N.; Houghton, W.J.; Hannaman, G.W.; Joksimovic, V.

1981-01-01

Probabilistic Risk Assessment methods have been applied to gas-cooled reactors for more than a decade and to HTGRs for more than six years in the programs sponsored by the U.S. Department of Energy. Significant advancements to the development of PRA methodology in these programs are summarized as are the specific applications of the methods to HTGRs. Emphasis here is on PRA as a tool for evaluating HTGR design options. Current work and future directions are also discussed. (author)

Dependent Human Error Probability Assessment

International Nuclear Information System (INIS)

Simic, Z.; Mikulicic, V.; Vukovic, I.

2006-01-01

This paper presents an assessment of the dependence between dynamic operator actions modeled in a Nuclear Power Plant (NPP) PRA and estimate the associated impact on Core damage frequency (CDF). This assessment was done improve HEP dependencies implementation inside existing PRA. All of the dynamic operator actions modeled in the NPP PRA are included in this assessment. Determining the level of HEP dependence and the associated influence on CDF are the major steps of this assessment. A decision on how to apply the results, i.e., should permanent HEP model changes be made, is based on the resulting relative CDF increase. Some CDF increase was selected as a threshold based on the NPP base CDF value and acceptance guidelines from the Regulatory Guide 1.174. HEP dependence resulting in a CDF increase of > 5E-07 would be considered potential candidates for specific incorporation into the baseline model. The approach used to judge the level of dependence between operator actions is based on dependency level categories and conditional probabilities developed in the Handbook of Human Reliability Analysis with Emphasis on Nuclear Power Plant Applications NUREG/CR-1278. To simplify the process, NUREG/CR-1278 identifies five levels of dependence: ZD (zero dependence), LD (low dependence), MD (moderate dependence), HD (high dependence), and CD (complete dependence). NUREG/CR-1278 also identifies several qualitative factors that could be involved in determining the level of dependence. Based on the NUREG/CR-1278 information, Time, Function, and Spatial attributes were judged to be the most important considerations when determining the level of dependence between operator actions within an accident sequence. These attributes were used to develop qualitative criteria (rules) that were used to judge the level of dependence (CD, HD, MD, LD, ZD) between the operator actions. After the level of dependence between the various HEPs is judged, quantitative values associated with the

A methodology for reviewing Probabilistic Risk Assessments

International Nuclear Information System (INIS)

Derby, S.L.

1983-01-01

The starting point for peer review of a Probabilistic Risk Assessment (PRA) is a clear understanding of how the risk estimate was prepared and of what contributions dominate the calculation. The problem facing the reviewers is how to cut through the complex details of a PRA to gain this understanding. This paper presents a structured, analytical procedure that solves this problem. The effectiveness of this solution is demonstrated by an application on the Zion Probabilistic Safety Study. The procedure found the three dominant initiating events and provided a simplified reconstruction of the calculation of the risk estimate. Significant assessments of uncertainty were also identified. If peer review disputes the accuracy of these judgments, then the revised risk estimate could significantly increase. The value of this procedure comes from having a systematic framework for the PRA review. Practical constraints limit the time and qualified people needed for an adequate review. Having the established framework from this procedure as a starting point, reviewers can focus most of their attention on the accuracy and the completeness of the calculation. Time wasted at the start of the review is reduced by first using this procedure to sort through the technical details of the PRA and to reconstruct the risk estimate from dominant contributions

Studi Awal Pra Desain Pabrik Bioetanol dari Nira Siwalan

Directory of Open Access Journals (Sweden)

Novarian Budisetyowati

2017-01-01

Full Text Available Bioetanol kini banyak dikembangkan sebagai bahan bakar alternatif pengganti bahan bakar fosil. Bioetanol untuk campuran bensin harus memiliki kemurnian sebesar 99,5-100%. Bioetanol dapat diperoleh dengan proses fermentasi yang melibatkan mikroorganisme. Pra desain pabrik bioetanol dari nira siwalan ini menggunakan proses fermentasi. Bahan baku berupa nira siwalan diasamkan dengan menggunakan H2SO4, kemudian disterilisasi sebelum difermentasi di fermentor selama 36 jam. Adapun mikroorganisme yang digunakan adalah Saccharomyces cereviceae. Bakteri ini mampu mengurai gula tanpa kehadiran oksigen dan menghasilkan etanol dan karbondioksida. Bioetanol dapat diperoleh dengan proses fermentasi yang melibatkan mikroorganisme. Pra desain pabrik bioetanol dari nira siwalan ini menggunakan proses fermentasi. Bahan baku berupa nira siwalan diasamkan dengan menggunakan H2SO4, kemudian disterilisasi sebelum difermentasi di fermentor selama 36 jam. Adapun mikroorganisme yang digunakan adalah Saccharomyces cereviceae. Setelah dari fermentor nira yang sudah difermentasi dinetralkan pH nya menggunakan NH4OH di tangki netralisasi. Dari tangki netralisasi nira dipompakan melewati preheater sebelum masuk ke kolom distilasi. Pemurnian dilakukan dengan menggunakan kolom distilasi sebanyak 2 buah. Pada distilasi yang pertama diperoleh kadar etanol sebesar 60% dan pada distilasi yang kedua diperoleh kadar 96%. Dari kolom distilasi 2 larutan didinginkan menggunakan cooler untuk didapatkan suhu 32oC agar sesuai dengan suhu proses dehidrasi dengan menggunakan Molecular Sieve yang diinginkan. Proses dehidrasi dilakukan untuk mendapat kadar etanol 99,5%. Etanol 99,5% yang dihasilkan kemudian disimpan dalam tangki penampung. Kebutuhan bioetanol dalam negeri pada tahun 2018 diperkirakan 3.166.015,13 kL/tahun. Berdasarkan analisa ekonomi yang dilakukan, diperoleh hasil sebagai berikut internal rate of return 26,53 % per tahun, pay out time 4,73 tahun, dan BEP 34,62 % Ditinjau

System Analysis and Risk Assessment (SARA) system

International Nuclear Information System (INIS)

Krantz, E.A.; Russell, K.D.; Stewart, H.D.; Van Siclen, V.S.

1986-01-01

Utilization of Probabilistic Risk Assessment (PRA) related information in the day-to-day operation of plant systems has, in the past, been impracticable due to the size of the computers needed to run PRA codes. This paper discusses a microcomputer-based database system which can greatly enhance the capability of operators or regulators to incorporate PRA methodologies into their routine decision making. This system is called the System Analysis and Risk Assessment (SARA) system. SARA was developed by EG and G Idaho, Inc. at the Idaho National Engineering Laboratory to facilitate the study of frequency and consequence analyses of accident sequences from a large number of light water reactors (LWRs) in this country. This information is being amassed by several studies sponsored by the United States Nuclear Regulatory Commission (USNRC). To meet the need of portability and accessibility, and to perform the variety of calculations necessary, it was felt that a microcomputer-based system would be most suitable

Introduction of accidental procedures in the event trees of the 900MW PWR PRA

International Nuclear Information System (INIS)

Bars, G.; Champ, M.; Lanore, J.M.; Pochard, R.

1985-02-01

This paper presents the example of the small LOCA Event Trees and the studies related to the introduction of procedure actions is case of HPSI failure. The results illustrate the interest of the approach and its significant impact on the PRA. The present studies are related to the Y actions in case of small LOCAs without HPIS

Spatial interactions database development for effective probabilistic risk assessment

International Nuclear Information System (INIS)

Liming, J. K.; Dunn, R. F.

2008-01-01

In preparation for a subsequent probabilistic risk assessment (PRA) fire risk analysis update, the STP Nuclear Operating Company (STPNOC) is updating its spatial interactions database (SID). This work is being performed to support updating the spatial interactions analysis (SIA) initially performed for the original South Texas Project Electric Generating Station (STPEGS) probabilistic safely assessment (PSA) and updated in the STPEGS Level 2 PSA and IPE Report. S/A is a large-scope screening analysis performed for nuclear power plant PRA that serves as a prerequisite basis for more detailed location-dependent, hazard-spec analyses in the PRA, such as fire risk analysis, flooding risk analysis, etc. SIA is required to support the 'completeness' argument for the PRA scope. The objectives of the current SID development effort are to update the spatial interactions analysis data, to the greatest degree practical, to be consistent with the following: the as-built plant as of December 31, 2007 the in-effect STPNOC STPEGS Units 1 and 2 PRA the current technology and intent of NUREG/CR-6850 guidance for lire risk analysis database support the requirements for PRA SIA, including fire and flooding risk analysis, established by NRC Regulatory Guide 1.200 and the ASME PRA Standard (ASME RA-S-2002 updated through ASME RA-Sc-2007,) This paper presents the approach and methodology for state-of-the-art SID development and applications, including an overview of the SIA process for nuclear power plant PRA. The paper shows how current relational database technology and existing, conventional station information sources can be employed to collect, process, and analyze spatial interactions data for the plant in an effective and efficient manner to meet the often challenging requirements of industry guidelines and standards such as NUREG/CR-6850, NRC Regulatory Guide 1.200, and ASME RA-S-2002 (updated through ASME RA-Sc 2007). This paper includes tables and figures illustrating how SIA

Method and system for dynamic probabilistic risk assessment

Science.gov (United States)

Dugan, Joanne Bechta (Inventor); Xu, Hong (Inventor)

2013-01-01

The DEFT methodology, system and computer readable medium extends the applicability of the PRA (Probabilistic Risk Assessment) methodology to computer-based systems, by allowing DFT (Dynamic Fault Tree) nodes as pivot nodes in the Event Tree (ET) model. DEFT includes a mathematical model and solution algorithm, supports all common PRA analysis functions and cutsets. Additional capabilities enabled by the DFT include modularization, phased mission analysis, sequence dependencies, and imperfect coverage.

A technique for human error analysis (ATHEANA)

International Nuclear Information System (INIS)

Cooper, S.E.; Ramey-Smith, A.M.; Wreathall, J.; Parry, G.W.

1996-05-01

Probabilistic risk assessment (PRA) has become an important tool in the nuclear power industry, both for the Nuclear Regulatory Commission (NRC) and the operating utilities. Human reliability analysis (HRA) is a critical element of PRA; however, limitations in the analysis of human actions in PRAs have long been recognized as a constraint when using PRA. A multidisciplinary HRA framework has been developed with the objective of providing a structured approach for analyzing operating experience and understanding nuclear plant safety, human error, and the underlying factors that affect them. The concepts of the framework have matured into a rudimentary working HRA method. A trial application of the method has demonstrated that it is possible to identify potentially significant human failure events from actual operating experience which are not generally included in current PRAs, as well as to identify associated performance shaping factors and plant conditions that have an observable impact on the frequency of core damage. A general process was developed, albeit in preliminary form, that addresses the iterative steps of defining human failure events and estimating their probabilities using search schemes. Additionally, a knowledge- base was developed which describes the links between performance shaping factors and resulting unsafe actions

A technique for human error analysis (ATHEANA)

Energy Technology Data Exchange (ETDEWEB)

Cooper, S.E.; Ramey-Smith, A.M.; Wreathall, J.; Parry, G.W. [and others

1996-05-01

Probabilistic risk assessment (PRA) has become an important tool in the nuclear power industry, both for the Nuclear Regulatory Commission (NRC) and the operating utilities. Human reliability analysis (HRA) is a critical element of PRA; however, limitations in the analysis of human actions in PRAs have long been recognized as a constraint when using PRA. A multidisciplinary HRA framework has been developed with the objective of providing a structured approach for analyzing operating experience and understanding nuclear plant safety, human error, and the underlying factors that affect them. The concepts of the framework have matured into a rudimentary working HRA method. A trial application of the method has demonstrated that it is possible to identify potentially significant human failure events from actual operating experience which are not generally included in current PRAs, as well as to identify associated performance shaping factors and plant conditions that have an observable impact on the frequency of core damage. A general process was developed, albeit in preliminary form, that addresses the iterative steps of defining human failure events and estimating their probabilities using search schemes. Additionally, a knowledge- base was developed which describes the links between performance shaping factors and resulting unsafe actions.

Developing and evaluating distributions for probabilistic human exposure assessments

Energy Technology Data Exchange (ETDEWEB)

Maddalena, Randy L.; McKone, Thomas E.

2002-08-01

This report describes research carried out at the Lawrence Berkeley National Laboratory (LBNL) to assist the U. S. Environmental Protection Agency (EPA) in developing a consistent yet flexible approach for evaluating the inputs to probabilistic risk assessments. The U.S. EPA Office of Emergency and Remedial Response (OERR) recently released Volume 3 Part A of Risk Assessment Guidance for Superfund (RAGS), as an update to the existing two-volume set of RAGS. The update provides policy and technical guidance on performing probabilistic risk assessment (PRA). Consequently, EPA risk managers and decision-makers need to review and evaluate the adequacy of PRAs for supporting regulatory decisions. A critical part of evaluating a PRA is the problem of evaluating or judging the adequacy of input distributions PRA. Although the overarching theme of this report is the need to improve the ease and consistency of the regulatory review process, the specific objectives are presented in two parts. The objective of Part 1 is to develop a consistent yet flexible process for evaluating distributions in a PRA by identifying the critical attributes of an exposure factor distribution and discussing how these attributes relate to the task-specific adequacy of the input. This objective is carried out with emphasis on the perspective of a risk manager or decision-maker. The proposed evaluation procedure provides consistency to the review process without a loss of flexibility. As a result, the approach described in Part 1 provides an opportunity to apply a single review framework for all EPA regions and yet provide the regional risk manager with the flexibility to deal with site- and case-specific issues in the PRA process. However, as the number of inputs to a PRA increases, so does the complexity of the process for calculating, communicating and managing risk. As a result, there is increasing effort required of both the risk professionals performing the analysis and the risk manager

Variáveis meteorológicas e cobertura vegetal de espécies arbóreas em praças urbanas em Cuiabá, Brasil

Directory of Open Access Journals (Sweden)

Angela Santana de Oliveira

2013-12-01

Full Text Available A influência da vegetação nas variáveis meteorológicas foi avaliada por meio do índice de área foliar (IAF e índice de sombreamento arbóreo (ISA em duas praças públicas em Cuiabá-MT, Brasil. Medidas de temperatura do ar (T e umidade relativa (UR foram obtidas sob a copa das árvores em diferentes sítios da cidade para o período seco e chuvoso no ano de 2009. A análise dos valores médios destas variáveis mostraram maiores valores de T e menores UR ocorrendo durante o período seco e sendo semelhantes nas duas praças. Com relação à UR, entretanto, não houve diferenças significativas entre a medida sob as árvores e a atmosfera. O índice de área foliar foi calculado e variou em função das espécies arbóreas das praças, e mostrou valores entre 5,64 e 2,79 m². m-2, sendo a média do IAF e do ISA na Praça Popular superiores ao da Praça 8 de Abril. Conclui-se que as espécies arbóreas melhoraram o ambiente térmico em virtude da atenuação da radiação proporcionada pelo sombreamento das diferentes espécies, principalmente no horário com menor ângulo solar.

Risk-informed design of IRIS using a level-1 probabilistic risk assessment from its conceptual design phase

International Nuclear Information System (INIS)

Mizuno, Yuko; Ninokata, Hisashi; Finnicum, David J.

2005-01-01

In this study, a probabilistic risk assessment (PRA) for the International Reactor Innovative and Secure (IRIS) has been generated to address two key areas as a part of the effort for the pre-application licensing of the IRIS design. First, the IRIS PRA is supporting the evaluation of IRIS design by providing design insights as well as a solid risk basis for the pre-licensing evaluation of the IRIS design. Second, the current PRA task is beginning the preparation of the more complete PRA analyses and documentation that will be required for Design Certification. The initial IRIS PRA is an at-power, Level-1 PRA for internal events that focuses on the evaluation of the IRIS design features to support the risk-informed design of IRIS by application of the PRA insights and the risk information to the design. To accomplish the evaluation, a reasonably complete Level-1 PRA model has been developed. The use of PRA in the early stages of the design has allowed a selection of design and performance features and an optimization of the design of several systems to reduce the potential for events that could lead to core damage via both enhanced prevention and mitigation of challenges. As a result, the total core damage frequency for internal events for the IRIS design has been calculated as 1.2x10 -8 per year

Hybrid causal methodology and software platform for probabilistic risk assessment and safety monitoring of socio-technical systems

Energy Technology Data Exchange (ETDEWEB)

Groth, Katrina, E-mail: kgroth@umd.ed [Center for Risk and Reliability, 0151 Glenn L. Martin Hall, University of Maryland, College Park, MD 20742 (United States); Wang Chengdong; Mosleh, Ali [Center for Risk and Reliability, 0151 Glenn L. Martin Hall, University of Maryland, College Park, MD 20742 (United States)

2010-12-15

This paper introduces an integrated framework and software platform for probabilistic risk assessment (PRA) and safety monitoring of complex socio-technical systems. An overview of the three-layer hybrid causal logic (HCL) modeling approach and corresponding algorithms, implemented in the Trilith software platform, are provided. The HCL approach enhances typical PRA methods by quantitatively including the influence of soft causal factors introduced by human and organizational aspects of a system. The framework allows different modeling techniques to be used for different aspects of the socio-technical system. The HCL approach combines the power of traditional event sequence diagram (ESD)event tree (ET) and fault tree (FT) techniques for modeling deterministic causal paths, with the flexibility of Bayesian belief networks for modeling non-deterministic cause-effect relationships among system elements (suitable for modeling human and organizational influences). Trilith enables analysts to construct HCL models and perform quantitative risk assessment and management of complex systems. The risk management capabilities included are HCL-based risk importance measures, hazard identification and ranking, precursor analysis, safety indicator monitoring, and root cause analysis. This paper describes the capabilities of the Trilith platform and power of the HCL algorithm by use of example risk models for a type of aviation accident (aircraft taking off from the wrong runway).

Hybrid causal methodology and software platform for probabilistic risk assessment and safety monitoring of socio-technical systems

International Nuclear Information System (INIS)

Groth, Katrina; Wang Chengdong; Mosleh, Ali

2010-01-01

This paper introduces an integrated framework and software platform for probabilistic risk assessment (PRA) and safety monitoring of complex socio-technical systems. An overview of the three-layer hybrid causal logic (HCL) modeling approach and corresponding algorithms, implemented in the Trilith software platform, are provided. The HCL approach enhances typical PRA methods by quantitatively including the influence of soft causal factors introduced by human and organizational aspects of a system. The framework allows different modeling techniques to be used for different aspects of the socio-technical system. The HCL approach combines the power of traditional event sequence diagram (ESD)event tree (ET) and fault tree (FT) techniques for modeling deterministic causal paths, with the flexibility of Bayesian belief networks for modeling non-deterministic cause-effect relationships among system elements (suitable for modeling human and organizational influences). Trilith enables analysts to construct HCL models and perform quantitative risk assessment and management of complex systems. The risk management capabilities included are HCL-based risk importance measures, hazard identification and ranking, precursor analysis, safety indicator monitoring, and root cause analysis. This paper describes the capabilities of the Trilith platform and power of the HCL algorithm by use of example risk models for a type of aviation accident (aircraft taking off from the wrong runway).

Nuclear power plant personnel errors in decision-making as an object of probabilistic risk assessment

International Nuclear Information System (INIS)

Reer, B.

1993-09-01

The integration of human error - also called man-machine system analysis (MMSA) - is an essential part of probabilistic risk assessment (PRA). A new method is presented which allows for a systematic and comprehensive PRA inclusions of decision-based errors due to conflicts or similarities. For the error identification procedure, new question techniques are developed. These errors are shown to be identified by looking at retroactions caused by subordinate goals as components of the overall safety relevant goal. New quantification methods for estimating situation-specific probabilities are developed. The factors conflict and similarity are operationalized in a way that allows their quantification based on informations which are usually available in PRA. The quantification procedure uses extrapolations and interpolations based on a poor set of data related to decision-based errors. Moreover, for passive errors in decision-making a completely new approach is presented where errors are quantified via a delay initiating the required action rather than via error probabilities. The practicability of this dynamic approach is demonstrated by a probabilistic analysis of the actions required during the total loss of feedwater event at the Davis-Besse plant 1985. The extensions of the ''classical'' PRA method developed in this work are applied to a MMSA of the decay heat removal (DHR) of the ''HTR-500''. Errors in decision-making - as potential roots of extraneous acts - are taken into account in a comprehensive and systematic manner. Five additional errors are identified. However, the probabilistic quantification results a nonsignificant increase of the DHR failure probability. (orig.) [de

Análise quali-quantitativa da arborização na praça XV de novembro em Ribeirão Preto - SP, Brasil

Directory of Open Access Journals (Sweden)

Gustavo de Nobrega Romani

2012-06-01

Full Text Available A Praça XV de Novembro, implantada em meados do século XIX, tem grande valor histórico-cultural, além de se constituir em uma das principais áreas verdes do centro da cidade de Ribeirão Preto. Visando ao conhecimento detalhado da vegetação para fins de orientação do manejo e conservação dessa área, foi feito um levantamento quali-quantitativo e fitossociológico das árvores e palmeiras da praça. Foram medidas altura e Diâmetro à Altura do Peito (DAP e identificados todos os indivíduos de porte arbóreo (árvores e palmeiras presentes na Praça, em nível de espécie. A praça ocupa uma área de 15.456,00 m², onde foram amostradas 42 espécies distribuídas por 19 famílias, num total de 161 indivíduos. Apesar de o local apresentar arborização com alto índice de diversidade de espécies (Shannon-Weaver de 3,14, os exemplares necessitam de maior atenção quanto a problemas ligados à fitossanidade e podas adequadas, fazendo que resulte em espaço seguro para os frequentadores e em boa qualidade paisagística.

Failure rate data for fusion safety and risk assessment

International Nuclear Information System (INIS)

Cadwallader, L.C.

1993-01-01

The Fusion Safety Program (FSP) at the Idaho National Engineering Laboratory (INEL) conducts safety research in materials, chemical reactions, safety analysis, risk assessment, and in component research and development to support existing magnetic fusion experiments and also to promote safety in the design of future experiments. One of the areas of safety research is applying probabilistic risk assessment (PRA) methods to fusion experiments. To apply PRA, we need a fusion-relevant radiological dose code and a component failure rate data base. This paper describes the FSP effort to develop a failure rate data base for fusion-specific components

Nuclear Regulatory Commission probabilistic risk assessment implementation program: A status report

International Nuclear Information System (INIS)

Rubin, M.P.; Caruso, M.A.

1996-01-01

The US Nuclear Regulatory Commission (NRC) is undertaking a number of activities intended to increase the consideration of risk significance in its decision processes and the effective use of risk-based technologies in its regulatory activities. Although the NRC is moving toward risk-informed regulation throughout its areas of responsibilities, this paper focuses primarily on those issues associated with reactor regulation. As the NRC completed significant milestones in its development of probabilistic risk assessment (PRA) methodology and gained considerable experience in the limited application of risk assessment to selected regulatory activities, it became evident that a much broader use of risk informed approaches offered advantages to both the NRC and the US commercial nuclear industry. This desire to enhance the use of risk assessment is driven by the clear belief that application of PRA methods will result in direct improvements in nuclear power plant operational safety from the perspective of both the regulator and the plant operator. The NRC believed that an overall policy on the use of PRA methods in nuclear regulatory activities should be established so that the many potential applications of PRA could be implemented in a consistent and predictable manner that would promote regulatory stability and efficiency. This paper describes the key activities that the NRC has undertaken to implement the initial stages of an integrated risk-informed regulatory framework

Assessment of uncertainties in severe accident management strategies

International Nuclear Information System (INIS)

Kastenberg, W.E.; Apostolakis, G.; Catton, I.; Dhir, V.K.; Okrent, D.

1990-01-01

Recent progress on the development of Probabilistic Risk Assessment (PRA) as a tool for qualifying nuclear reactor safety and on research devoted to severe accident phenomena has made severe accident management an achievable goal. Severe accident management strategies may involve operational changes, modification and/or addition of hardware, and institutional changes. In order to achieve the goal of managing severe accidents, a method for assessment of strategies must be developed which integrates PRA methodology and our current knowledge concerning severe accident phenomena, including uncertainty. The research project presented in this paper is aimed at delineating uncertainties in severe accident progression and their impact on severe accident management strategies

Avaliação qualitativa e quantitativa da arborização das praças de Vinhedo, SP.

Directory of Open Access Journals (Sweden)

Roberval de Cássia Salvador Ribeiro

2006-06-01

Full Text Available O inventário das espécies arbóreas e dos respectivos números de indivíduos das praças da cidade de Vinhedo foi realizado no perímetro urbano, excetuando-se os condomínios, as áreas de parques e as de preservação de mananciais. Para a localização das áreas, consultou-se a planta do município de 1997. Realizou-se o inventário da vegetação arbórea, considerando-se apenas os indivíduos com CAP (circunferência à altura do peito acima de 10 cm listando-se as seguintes informações: nomes comum e científico das espécies; CAP; altura; aspecto geral; diâmetro de copa; presença de pragas, doenças ou parasitas; ocorrência de podas (drástica e/ou de condução; fitossanidade da raiz, tronco e copa. Foram registradas 22 praças por nome, localização e número total de árvores, totalizando 764 indivíduos pertencentes a 23 famílias botânicas e 53 espécies, além de 32 indivíduos não identificados. A espécie de maior abundância relativa foi Syagrus romanzoffiana (jerivá, com 31,94% do número total de indivíduos. Em 63,64% das praças 33,13% das espécies eram exóticas. A maior parte dos indivíduos tinha aspecto geral normal, demonstrando prática de tratos culturais adequados. Na maioria dos casos, as podas foram feitas corretamente, ou não houve a necessidade de nenhuma intervenção. Do total de 22 praças, apenas cinco tinham bom estado geral de conservação dos elementos naturais (arbustos, canteiros e gramados. Em 68,18% das praças as árvores tinham altura superior a 6 metros, indicando que essas áreas necessitavam apenas de procedimentos de manutenção de rotina. E 22,72% necessitavam de práticas de manutenção mais direcionadas ao desenvolvimento das árvores, tais como adubações periódicas, capinas, podas de condução e, finalmente, em 13,64% deveriam ocorrer intervenções tanto de manutenção, como de recuperação por meio de novos plantios, ou mesmo, de planejamento para remodelação da área.

Primer uticaja filtriranja slike u sistemima za praćenje ciljeva primenom termovizije / An example of image filtering in target tracking systems with thermal imagery

Directory of Open Access Journals (Sweden)

Zvonko M. Radosavljević

2003-07-01

Full Text Available U radu je dat primer primene jedne vrste niskofrekventnog filtriranja sa usrednjavanjem, koje se primenjuje u sistemima za detekciju i praćenje ciljeva u vazdušnom prostoru primenom termovizije. Date su dve metode filtriranja slike. Prva metoda koristi niskofrekventno konvoluciono filtriranje a druga usrednjavajući filtar na osnovu srednje vrednosti nivoa sivog. Ovi filtri su primenjeni u sistemima za praćenje uz pomoć infracrvenih senzora. Određivanje nivoa praga filtriranja vrši se uz pomoć statističkih osobina slike. Veoma važan korak u procesu praćenja je određivanje prozora praćenja, koji maze biti, po dimenzijama, fiksan ili adaptibilan. Pogrešna procena o postojanju cilja u prozoru može se doneti u slučaju prisustva šuma pozadine, predpojačavača, detektora, itd. Filtriranje je neophodan korak u ovim sistemima, kao značajan činilac U povećanju brzine i tačnosti praćenja. / A case of image filtering in air target detecting and tracking systems is described in this paper. Two image filtering methods are given. The first method is performed using a low pass convolving filter and the second one uses the mean value of gray level filter. The main goal of the cited filtering is implementation in IR (infra red systems. Some statistical features of the images were used for selecting the threshold level. The next step in the algorithm is the determination of a 'tracking window' that can be fixed or adaptive in size. A false estimation of a target existing in the window may be influenced by the background noise, low noise amplifier detector, etc.

Alternative Assessment Techniques.

Science.gov (United States)

Lowenthal, Barbara

1988-01-01

Maintaining the precision necessary for administering norm referenced tests can be a problem for the special education teacher who is trained to assist the student. Criterion-referenced tests, observations, and interviews are presented as effective alternative assessment techniques. (JDD)

Beyond informed choice: Prenatal risk assessment, decision-making and trust

Directory of Open Access Journals (Sweden)

Nete Schwennesen

2008-05-01

Full Text Available In 2004 prenatal risk assessment (PRA was implemented as a routine offer to all pregnant women in Denmark. It was argued that primarily the new programme would give all pregnant women an informed choice about whether to undergo prenatal testing. On the basis of ethnographic fieldwork in an ultrasound clinic in Denmark and interviews with pregnant women and their partners, we call into question the assumption underlying the new guidelines that more choice and more objective information is a source of empowerment and control. We focus on one couple's experience of PRA. This case makes it evident how supposed choices in the context of PRA may not be experienced as such. Rather, they are experienced as complicated processes of meaning-making in the relational space between the clinical setting, professional authority and the social life of the couples. PRA users are reluctant to make choices and abandon health professionals as authoritative experts in the face of complex risk knowledge. When assumptions about autonomy and self-determination are inscribed into the social practice of PRA, authority is transferred to the couple undergoing PRA and a new configuration of responsibility evolves between the couple and their relationship to the foetus. It is argued that al-though the new programme of prenatal testing in Denmark presents itself in opposition to quasi-eugenic and paternalistic forms of governing couples' decisions it represents another form of government that works through the notion of choice. An ethics of a shared responsibility of PRA and its outcome would be more in agreement with how decisions are actually made.http://dx.doi.org/10.5324/eip.v2i1.1687

Risk Management of NASA Projects

Science.gov (United States)

Sarper, Hueseyin

1997-01-01

Various NASA Langley Research Center and other center projects were attempted for analysis to obtain historical data comparing pre-phase A study and the final outcome for each project. This attempt, however, was abandoned once it became clear that very little documentation was available. Next, extensive literature search was conducted on the role of risk and reliability concepts in project management. Probabilistic risk assessment (PRA) techniques are being used with increasing regularity both in and outside of NASA. The value and the usage of PRA techniques were reviewed for large projects. It was found that both civilian and military branches of the space industry have traditionally refrained from using PRA, which was developed and expanded by nuclear industry. Although much has changed with the end of the cold war and the Challenger disaster, it was found that ingrained anti-PRA culture is hard to stop. Examples of skepticism against the use of risk management and assessment techniques were found both in the literature and in conversations with some technical staff. Program and project managers need to be convinced that the applicability and use of risk management and risk assessment techniques is much broader than just in the traditional safety-related areas of application. The time has come to begin to uniformly apply these techniques. The whole idea of risk-based system can maximize the 'return on investment' that the public demands. Also, it would be very useful if all project documents of NASA Langley Research Center, pre-phase A through final report, are carefully stored in a central repository preferably in electronic format.

Risk assessment under deep uncertainty: A methodological comparison

International Nuclear Information System (INIS)

Shortridge, Julie; Aven, Terje; Guikema, Seth

2017-01-01

Probabilistic Risk Assessment (PRA) has proven to be an invaluable tool for evaluating risks in complex engineered systems. However, there is increasing concern that PRA may not be adequate in situations with little underlying knowledge to support probabilistic representation of uncertainties. As analysts and policy makers turn their attention to deeply uncertain hazards such as climate change, a number of alternatives to traditional PRA have been proposed. This paper systematically compares three diverse approaches for risk analysis under deep uncertainty (qualitative uncertainty factors, probability bounds, and robust decision making) in terms of their representation of uncertain quantities, analytical output, and implications for risk management. A simple example problem is used to highlight differences in the way that each method relates to the traditional risk assessment process and fundamental issues associated with risk assessment and description. We find that the implications for decision making are not necessarily consistent between approaches, and that differences in the representation of uncertain quantities and analytical output suggest contexts in which each method may be most appropriate. Finally, each methodology demonstrates how risk assessment can inform decision making in deeply uncertain contexts, informing more effective responses to risk problems characterized by deep uncertainty. - Highlights: • We compare three diverse approaches to risk assessment under deep uncertainty. • A simple example problem highlights differences in analytical process and results. • Results demonstrate how methodological choices can impact risk assessment results.

Application of database management software to probabilistic risk assessment calculations

International Nuclear Information System (INIS)

Wyss, G.D.

1993-01-01

Probabilistic risk assessment (PRA) calculations require the management and processing of large amounts of information. This data normally falls into two general categories. For example, a commercial nuclear power plant PRA study makes use of plant blueprints and system schematics, formal plant safety analysis reports, incident reports, letters, memos, handwritten notes from plant visits, and even the analyst's ''engineering judgment''. This information must be documented and cross-referenced in order to properly execute and substantiate the models used in a PRA study. The first category is composed of raw data that is accumulated from equipment testing and operational experiences. These data describe the equipment, its service or testing conditions, its failure mode, and its performance history. The second category is composed of statistical distributions. These distributions can represent probabilities, frequencies, or values of important parameters that are not time-related. Probability and frequency distributions are often obtained by fitting raw data to an appropriate statistical distribution. Database management software is used to store both types of data so that it can be readily queried, manipulated, and archived. This paper provides an overview of the information models used for storing PRA data and illustrates the implementation of these models using examples from current PRA software packages

Perbandingan Tinggi Tulang Maksila dan Mandibula di Regio Interisisivi Sentral antara Pra dan Pasca Perawatan Ortodontik dengan Pencabutan ke Empat Gigi Premolar Pertama (Kajian pada Foto Panoramik

Directory of Open Access Journals (Sweden)

Wayan Ardhana

2012-12-01

Full Text Available Latar belakang. Perawatan ortodontik pada kasus-kasus gigi berjejal dan protusif sering membutuhkan pencabutan gigi premolar untuk penyediaan ruang agar gigi berjejal dapat dirapikan dan gigi depan yang protusif dapat diundurkan. Gigi insisivus sentral merupakan salah satu gigi yang paling banyak mengalami pergerakan selama proses retrusi. Pergerakan gigi insisivus mengakibatkan terjadinya perubahan pada puncak tulang alveolar selama perawatan yang mungkin akan mempengaruhi tinggi tulang maksila dan mandibula pasca perawatan. Tujuan penelitian. Membandingkan tinggi tulang maksila dan mandibula di daerah interdental gigi insisivi sentral pada foto panoramic antara pra dan pasca perawatan maloklusi dengan pencabutan ke empat gigi premolar pertama. Metode penelitian. Digunakan 30 pasang foto panoramic pra dan pasca perawatan yang dipilih sesuai dengan kriteria penelitian dari pasien-pasien peneliti yang telah selesai mendapat perawatan aktif dengan teknik edgewise. Analisis Kolmogorov-Smirnov dan Shaviro-Wilk digunakan untuk uji normalitas dan Student t-test data berpasangan digunakan untuk menguji perbedaan tinggi tulang maksila dan mandibula antara pra dan pasca perawatan. Hasil Penelitian. Tidak didapatkan perbedaan (p>0,05 tinggi tulang maksila dan amndibula antara pra dan pasca perawatan ortodontik dengan pencabutan keempat gigi premolar pertama.   Background. In orthodontic treatment, premolar extractions are often needed in crowding and prostrusive cases to provide space for the teeth can be aligned and retracted to their desire position. Central incisor teeth are the teeth that mostly undergone more movement during retrusion. The change of the alveolar bone crest in this incisors might affect the maxillary and mandibular bone height post-treatment. Research objectives. The present study aimed to compare the bone height in the interdental maxillary and mandibular central incisors regions before and after orthodontic treatment with four

Advanced Test Reactor probabilistic risk assessment

International Nuclear Information System (INIS)

Atkinson, S.A.; Eide, S.A.; Khericha, S.T.; Thatcher, T.A.

1993-01-01

This report discusses Level 1 probabilistic risk assessment (PRA) incorporating a full-scope external events analysis which has been completed for the Advanced Test Reactor (ATR) located at the Idaho National Engineering Laboratory

A methodology for reviewing probabilistic risk assessments

International Nuclear Information System (INIS)

Derby, S.L.

1983-01-01

The starting point for peer review of a Probabilistic Risk Assessment (PRA) is a clear understanding of how the risk estimate was prepared and of what contributions dominate the calculation. The problem facing the reviewers is how to cut through the complex details of a PRA to gain this understanding. This paper presents a structured, analytical procedure that solves this problem. The effectiveness of this solution is demonstrated by an application on the Zion Probabilistic Safety Study. The procedure found the three dominant initiating events and provided a simplified reconstruction of the calculation of the risk estimate. Significant assessments of uncertainty were also identified. If peer review disputes the accuracy of these judgments, then the revised risk estimate could significantly increase

Probabilistic risk assessment for salt repository conceptual design of subsurface facilities: A techical basis for Q-list determination

International Nuclear Information System (INIS)

Chen, C.P.; Mayberry, J.J.; Shepherd, J.; Koza, H.; Rahmani, H.; Sinsky, J.

1987-12-01

Subpart G ''Quality Assurance'' of 10 CFR Part 60 requires that the US Department of Energy (DOE) apply a quality assurance program to ''all systems, structures, and components important to safety'' and to ''design and characterization of barriers important to waste isolation.'' In April 1986, DOE's Office of Geologic Repositories (OGR) issued general guidance for formulating a list of such systems, structures, and components---the Q-list. This guidance called for the use of probabilistic risk assessment (PRA) techniques to identify Q-list items. In this report, PRA techniques are applied to the underground facilities and systems described in the conceptual design report for the Salt Repository Project (SRP) in Deaf Smith County, Texas. Based on probability and dose consequence calculations, no specific items were identified for the Q-list. However, evaluation of the analyses indicated that two functions are important in precluding off-site releases of radioactivity: disposal container integrity; and isolation of the underground facility by the heating, ventilation, and air conditioning (HVAC) systems. Items related to these functions are recommended for further evaluation as the repository design progresses. 13 refs., 20 figs

Case study on the use of PSA methods: Backfitting decisions

International Nuclear Information System (INIS)

1991-04-01

This case study illustrates the process of using probabilistic risk assessment (PRA) method to evaluate proposed backfits of nuclear power plants (NPP), which are intended to enhance the plant safety by improving equipment operability. Some examples of situations in which PRA techniques have been used to address backfit issues at operating NPPs are summarized. 2 refs, 5 figs, 4 tabs

Treatment of complementary events in event trees in constructing linked fault trees for level 1 and level 2 PRA

International Nuclear Information System (INIS)

Jo, Y. G.

2008-01-01

Complementary events in the event trees for a PRA model should be treated properly in order to evaluate plant risk correctly. In this study, the characteristics of the following three different cut-set generation methods were investigated first in order to find the best practical way for treating complementary events: 1) exact method which treats complementary events logically, 2) no-delete term method which does not treat complementary events at all, and 3) delete term method which treats complementary events by deleting nonsense cut-sets which are generated as a result of ignoring complementary events. Then, practical methods for treating complementary events in constructing linked fault trees for level 1 and level 2 PRA in EPRI R and R workstation software environment, where CAFTA is the fault tree editor and FORTE is the cut-set engine, were suggested and demonstrated. The suggested methods deal with the following selected four typical cases: Case 1: an event tree event (E) is represented by a fault tree gate whose inputs consist of only fault tree gates, Case 2: E is represented by a single basic event, Case 3: E is represented by an OR fault tree gate which has a single basic event and a fault tree gate as inputs, and Case 4: E is represented by an AND fault tree gate which has a single basic event and a fault tree gate as inputs. In the suggested methods, first the high level logic structures of event tree events are examined and restructured, if needed. Then, the delete term method, the exact method, and the combination of the two methods are applied to Case 1, Case 2, and Cases 3 and 4, respectively. Also, it is recommended to treat complementary events, using the suggested methods, before level 1 and level 2 PRA fault trees are coupled. It should be noted that the selected four typical cases may not cover all different cases encountered in level 1 and level 2 PRA modeling. However, a process similar to the one suggested in this study may be used to find

Risk assessment of a fusion-reactor fuel-processing system

International Nuclear Information System (INIS)

Bruske, S.Z.; Holland, D.F.

1983-07-01

The probabilistic risk assessment (PRA) methodology provides a means to systematically examine the potential for accidents that may result in a release of hazardous materials. This report presents the PRA for a typical fusion reactor fuel processing system. The system used in the analysis is based on the Tritium Systems Test Assembly, which is being operated at the Los Alamos National Laboratory. The results of the evaluation are presented in a probability-consequence plot that describes the probability of various accidental tritium release magnitudes

PRA-1 offshore platform start-up within seven days; Operacionalizacao da plataforma offshore PRA-1 em sete dias

Energy Technology Data Exchange (ETDEWEB)

Barbosa, Fernando; Mitidieri, Jorge; Faria, Jose Luis Coutinho de; Ribeiro, Juan Carlos; Moura, Mario Arthur [Construtora Norberto Oderbrecht S.A., Rio de Janeiro, RJ (Brazil)

2008-07-01

The technologic innovations are very hard features with regards to Offshore Engineering and Construction over the worldwide. The innovations only make sense since they are focus on the high productivity, safe job and cost reduction compared with the current technologies. Inside the scenario mentioned above is Construtora Norberto Odebrecht S.A. concept for the PRA-1 platform Engineering and Construction. Through a very advanced and innovation concept, it was defined as the Main Strategic Planning of the undertaking not use a temporary platform support (named in Brazil as 'Flotel') during the 'Hook-up', commissioning and star-up offshore phase. The success of the strategic made possible through the implementation of new engineering tools, and, besides this, through a very careful offshore planning focused on minimizing and make easier as much as possible the offshore activities. The planning can be basically spitted on the following parts: A- Onshore preparations (Assembly, Integration and Commissioning of the Utilities and Accommodation Modules) B- Offshore detailed planning of the critical activities concerning the start-up of the systems responsible for leaving the platform ready for 'live'. This operation was defined as 'seven days of platform live support' (main target of this paper). (author)

Space shuttle main propulsion pressurization system probabilistic risk assessment

International Nuclear Information System (INIS)

Plastiras, J.K.

1989-01-01

This paper reports that, in post-Challenger discussions with Congressional Committees and the National Research Council Risk Management Oversight Panel, criticism was levied against NASA because of the inability to prioritize the 1300+ single point failures. In the absence of a ranking it was difficult to determine where special effort was needed in failure evaluation, in design improvement, in management review of problems, and in flight readiness reviews. The belief was that the management system was overwhelmed by the quantity of critical hardware items that were on the Critical Items List (CIL) and that insufficient attention was paid to the items that required it. Congressional staff members from Congressman Markey's committee who have oversight responsibilities in the nuclear industry, and specifically over the nuclear power supplies for NASA's Galileo and Ulysses missions, felt very strongly that the addition of Probabilistic Risk Assessment (PRA) to the existing Failure Mode Effects Analysis/Hazard Analysis (FMEA/HA) methods was exceedingly important. Specifically, the Markey committee recognized that the inductive, qualitative component-oriented FMEA could be supplemented by the deductive, quantitative systems-oriented PRA. Furthermore, they felt that the PRA approach had matured to the extent that it could be used to assess risk, even with limited shuttle-specific failure data. NASA responded with arguments that the FMEA/HA had illuminated all significant failure modes satisfactorily and that no failure rate data base was available to support the PRA approach

Giving Student Groups a Stronger Voice: Using Participatory Research and Action (PRA) to Initiate Change to a Curriculum

Science.gov (United States)

O'Neill, Geraldine; McMahon, Sinead

2012-01-01

Traditional student feedback mechanisms have been criticised for being teacher-centred in design and, in particular, for their absence of transparent follow-up actions. In contrast, this study describes the process and the evaluation of a participatory research and action (PRA) approach used in an undergraduate physiotherapy degree. This approach…

Discrete dynamic event tree modeling and analysis of nuclear power plant crews for safety assessment

International Nuclear Information System (INIS)

Mercurio, D.

2011-01-01

Current Probabilistic Risk Assessment (PRA) and Human Reliability Analysis (HRA) methodologies model the evolution of accident sequences in Nuclear Power Plants (NPPs) mainly based on Logic Trees. The evolution of these sequences is a result of the interactions between the crew and plant; in current PRA methodologies, simplified models of these complex interactions are used. In this study, the Accident Dynamic Simulator (ADS), a modeling framework based on the Discrete Dynamic Event Tree (DDET), has been used for the simulation of crew-plant interactions during potential accident scenarios in NPPs. In addition, an operator/crew model has been developed to treat the response of the crew to the plant. The 'crew model' is made up of three operators whose behavior is guided by a set of rules-of-behavior (which represents the knowledge and training of the operators) coupled with written and mental procedures. In addition, an approach for addressing the crew timing variability in DDETs has been developed and implemented based on a set of HRA data from a simulator study. Finally, grouping techniques were developed and applied to the analysis of the scenarios generated by the crew-plant simulation. These techniques support the post-simulation analysis by grouping similar accident sequences, identifying the key contributing events, and quantifying the conditional probability of the groups. These techniques are used to characterize the context of the crew actions in order to obtain insights for HRA. The model has been applied for the analysis of a Small Loss Of Coolant Accident (SLOCA) event for a Pressurized Water Reactor (PWR). The simulation results support an improved characterization of the performance conditions or context of operator actions, which can be used in an HRA, in the analysis of the reliability of the actions. By providing information on the evolution of system indications, dynamic of cues, crew timing in performing procedure steps, situation

Proposal of methodology of tsunami accident sequence analysis induced by earthquake using DQFM methodology

International Nuclear Information System (INIS)

Muta, Hitoshi; Muramatsu, Ken

2017-01-01

Since the Fukushima-Daiichi nuclear power station accident, the Japanese regulatory body has improved and upgraded the regulation of nuclear power plants, and continuous effort is required to enhance risk management in the mid- to long term. Earthquakes and tsunamis are considered as the most important risks, and the establishment of probabilistic risk assessment (PRA) methodologies for these events is a major issue of current PRA. The Nuclear Regulation Authority (NRA) addressed the PRA methodology for tsunamis induced by earthquakes, which is one of the methodologies that should be enhanced step by step for the improvement and maturity of PRA techniques. The AESJ standard for the procedure of seismic PRA for nuclear power plants in 2015 provides the basic concept of the methodology; however, details of the application to the actual plant PRA model have not been sufficiently provided. This study proposes a detailed PRA methodology for tsunamis induced by earthquakes using the DQFM methodology, which contributes to improving the safety of nuclear power plants. Furthermore, this study also states the issues which need more research. (author)

Guidelines on the scope, content, and use of comprehensive risk assessment in the management of high-level nuclear waste transportation

International Nuclear Information System (INIS)

Golding, D.; White, A.

1990-12-01

This report discusses the scope of risk assessment strategies in the management of the transport of high-level radioactive wastes. In spite of the shortcomings of probabilistic risk assessment(PRA), the Transportation Needs Assessment recommended this as the preferred methodology to assess the risks of high level nuclear waste (HLNW) transportation. A PRA also will need to heed the lessons learned from the development and application of PRA elsewhere, such as in the nuclear power industry. A set of guidelines will aid this endeavor by outlining the appropriate scope, content, and use of a risk assessment which is more responsive to the uncertainties, human-technical interactions, social forces, and iterative relationship with risk management strategies, than traditional PRAS. This more expansive definition, which encompasses but is not totally reliant on rigorous data requirements and quantitative probability estimates, we term Comprehensive Risk Assessment (CRA) Guidelines will be developed in three areas: the limitations of existing methodologies and suggested modifications; CRA as part of a flexible, effective, adaptive risk management system for HLNW transportation; and, the use of CRA in risk communication

Top event prevention analysis: A deterministic use of PRA

International Nuclear Information System (INIS)

Worrell, R.B.; Blanchard, D.P.

1996-01-01

This paper describes the application of Top Event Prevention Analysis. The analysis finds prevention sets which are combinations of basic events that can prevent the occurrence of a fault tree top event such as core damage. The problem analyzed in this application is that of choosing a subset of Motor-Operated Valves (MOVs) for testing under the Generic Letter 89-10 program such that the desired level of safety is achieved while providing economic relief from the burden of testing all safety-related valves. A brief summary of the method is given, and the process used to produce a core damage expression from Level 1 PRA models for a PWR is described. The analysis provides an alternative to the use of importance measures for finding the important combination of events in a core damage expression. This application of Top Event Prevention Analysis to the MOV problem was achieve with currently available software

Espaços livres em praças inglesas do século XVII: o caso de Convent Garden

Directory of Open Access Journals (Sweden)

Maria Fernanda Derntl

2006-12-01

Full Text Available Covent Garden é considerada uma experiência modelar entre as primeiras transformações no espaço urbano ainda de aparência medieval de Londres no século XVII. A partir da iconografia da época e das indicações da bibliografia pertinente, procura-se apresentar uma análise abrangente da configuração formal dessa praça. Destaca-se a existência de diferentes tipos de espaços livres contribuindo para determinar seu desenho: além do espaço central, também fizeram parte da praça o pátio da igreja de St. Paul, a oeste e os jardins da mansão Bedford, ao sul. Conclui-se que o desenho apresentado por Covent Garden não era completamente regular ou simétrico e teria sido bem diferente das squares londrinas dos séculos XVII e XVIII.

Elastin and Mechanics of Pig Pericardial Resistance Arteries (pPRA)

DEFF Research Database (Denmark)

Bloksgaard, Maria; Leurgans, Thomas; Rosenstand, Kristoffer

Resistance arteries are remodeled in hypertension and diabetes. Elastin was reported to play a role herein. The parietal pericardium is opened during cardio-thoracic surgeries and might be a valuable biopsy for research in cardio-vascular diseases. We tested the hypothesis that resistance arteries...... can be isolated from the pericardium to study the micro-architecture of elastin and vascular wall mechanics. The pericardium of pigs served to test the hypothesis. pPRAs were microdissected. Their structure was examined using multiphoton excitation fluorescence microscopy. Diameter......-tension and pressure-diameter-length relationships were recorded in myographs. Findings are compared to rodent mesenteric resistance arteries and –basilar arteries (rMRA, rBA) with comparable lumen diameter (±300µm at 100mmHg). pPRA have no clear external elastic lamina (present in rMRA, but not rBA), scant elastin...

The development by means of trial assessments, of a procedure for radiological risk assessment of underground disposal of low and intermediate level radioactive wastes

International Nuclear Information System (INIS)

Thompson, B.G.J.

1987-01-01

Seven trials are outlined showing the development and testing of a procedure based upon pra using Monte Carlo simulators, to assess post closure risks from the underground disposal of low and intermediate level radioactive wastes. The PRA method is found to be more justifiable than the use of 'best estimates'. Problems of accounting for long-term environmental changes and of future human intrusions are discussed. The importance of achieving statistical convergence within practical time scales and resources and of accounting for the influence of different sources of systematic bias is emphasised. (orig.)

Architecture for Integrated Medical Model Dynamic Probabilistic Risk Assessment

Science.gov (United States)

Jaworske, D. A.; Myers, J. G.; Goodenow, D.; Young, M.; Arellano, J. D.

2016-01-01

Probabilistic Risk Assessment (PRA) is a modeling tool used to predict potential outcomes of a complex system based on a statistical understanding of many initiating events. Utilizing a Monte Carlo method, thousands of instances of the model are considered and outcomes are collected. PRA is considered static, utilizing probabilities alone to calculate outcomes. Dynamic Probabilistic Risk Assessment (dPRA) is an advanced concept where modeling predicts the outcomes of a complex system based not only on the probabilities of many initiating events, but also on a progression of dependencies brought about by progressing down a time line. Events are placed in a single time line, adding each event to a queue, as managed by a planner. Progression down the time line is guided by rules, as managed by a scheduler. The recently developed Integrated Medical Model (IMM) summarizes astronaut health as governed by the probabilities of medical events and mitigation strategies. Managing the software architecture process provides a systematic means of creating, documenting, and communicating a software design early in the development process. The software architecture process begins with establishing requirements and the design is then derived from the requirements.

Review of UCN 5,6 Fire PSA Model based on ANS Fire PRA Standard

International Nuclear Information System (INIS)

Yang, Joon Eon; Lee, Yoon Hwan

2006-12-01

Recently, under the de-regulation environment, nuclear industry has attempted various approaches to improve the economics of Nuclear Power Plants (NPP). This approach uses the fire risk and performance information to manage the resources effectively and efficiently that are used in the operation of NPP. In fire risk informed/performance-based decision/operation, fire PSA quality is one of the most important things. The nuclear industry and regulatory body of U.S.A have developed a measure to evaluate the quality of fire PSA. ANS (American Nuclear Society) has developed a guidance called 'ANS Fire PRA Methodology Standard'. However, in Korea, there have been no attempts to evaluate the quality of fire PSA model itself. Therefore, we cannot be sure about the quality of fire PSA whether or not the present fire PSA model can be used for the risk-informed applications such as mentioned above. We can say that the evaluation of fire PSA model quality is the basis for the fire risk informed/performance-based decision/operation. In this report, we have evaluated the quality of fire PSA model for Ulchin 5 and 6 units based on the ANS Fire PRA Standard. We, also, have derived what items are to be improved to upgrade the quality of fire PSA model and how it can be improved. This report can be used as the base of the fire risk informed/performance-based decision/operation work in Korea

Application of probabilistic risk assessment methodology to fusion

International Nuclear Information System (INIS)

Piet, S.J.

1985-07-01

Probabilistic Risk Assessment (PRA) tools are applied to general fusion issues in a systematic way, generally qualitatively. The potential value of PRA to general fusion safety and economic issues is discussed. Several important design insights result: possible fault interactions must be minimized (decouple fault conditions), inherently safe designs must include provision for passively handling loss of site power and loss of coolant conditions, the reliability of the vacuum boundary appears vital to maximizing facility availabilty and minimizing safety risk, and economic analyses appear to be incomplete without consideration of potential availability loss from forced outrages. A modification to PRA formalism is introduced, called the fault interaction matrix. The fault interaction matrix contains information concerning what initial fault condition could lead to other fault conditions and with what frequency. Thus, the fault interaction matrix represents a way to present and measure the degree to which a designer has decoupled possible fault conditions in his design

Advanced neutron source reactor probabilistic flow blockage assessment

International Nuclear Information System (INIS)

Ramsey, C.T.

1995-08-01

The Phase I Level I Probabilistic Risk Assessment (PRA) of the conceptual design of the Advanced Neutron Source (ANS) Reactor identified core flow blockage as the most likely internal event leading to fuel damage. The flow blockage event frequency used in the original ANS PRA was based primarily on the flow blockage work done for the High Flux Isotope Reactor (HFIR) PRA. This report examines potential flow blockage scenarios and calculates an estimate of the likelihood of debris-induced fuel damage. The bulk of the report is based specifically on the conceptual design of ANS with a 93%-enriched, two-element core; insights to the impact of the proposed three-element core are examined in Sect. 5. In addition to providing a probability (uncertainty) distribution for the likelihood of core flow blockage, this ongoing effort will serve to indicate potential areas of concern to be focused on in the preliminary design for elimination or mitigation. It will also serve as a loose-parts management tool

Probabilistic risk assessment support of emergency preparedness at the Savannah River Site

International Nuclear Information System (INIS)

O'Kula, K.R.; Baker, W.H.; Simpkins, A.A.; Taylor, R.P.; Wagner, K.C.; Amos, C.N.

1992-01-01

Integration of the Probabilistic Risk Assessment (PRA) for K Reactor operation into related technical areas at the Savannah River Site (SRS) includes coordination with several onsite organizations responsible for maintaining and upgrading emergency preparedness capabilities. Major functional categories of the PRA application are scenario development and source term algorithm enhancement. Insights and technologies from the SRS PRA have facilitated development of: (1) credible timelines for scenarios; (2) algorithms tied to plant instrumentation to provide best-estimate source terms for dose projection; and (3) expert-system logic models to implement informed counter-measures to assure onsite and offsite safety following accidental releases. The latter methodology, in particular, is readily transferable to other reactor and non-reactor facilities at SRS and represents a distinct advance relative to emergency preparedness capabilities elsewhere in the DOE complex

Implementation of condition-dependent probabilistic risk assessment using surveillance data on passive components

International Nuclear Information System (INIS)

Lewandowski, Radoslaw; Denning, Richard; Aldemir, Tunc; Zhang, Jinsuo

2016-01-01

Highlights: • Condition-dependent probabilistic risk assessment (PRA). • Time-dependent characterization of plant-specific risk. • Containment bypass involving in secondary system piping and SCC in SG tubes. - Abstract: A great deal of surveillance data are collected for a nuclear power plant that reflect the changing condition of the plant as it ages. Although surveillance data are used to determine failure probabilities of active components for the plant’s probabilistic risk assessment (PRA) and to indicate the need for maintenance activities, they are not used in a structured manner to characterize the evolving risk of the plant. The present study explores the feasibility of using a condition-dependent PRA framework that takes a first principles approach to modeling the progression of degradation mechanisms to characterize evolving risk, periodically adapting the model to account for surveillance results. A case study is described involving a potential containment bypass accident sequence due to the progression of flow-accelerated corrosion in secondary system piping and stress corrosion cracking of steam generator tubes. In this sequence, a steam line break accompanied by failure to close of a main steam isolation valve results in depressurization of the steam generator and induces the rupture of one or more faulted steam generator tubes. The case study indicates that a condition-dependent PRA framework might be capable of providing early identification of degradation mechanisms important to plant risk.

Application of sensitivity analysis in nuclear power plant probabilistic risk assessment studies

International Nuclear Information System (INIS)

Hirschberg, S.; Knochenhauer, M.

1986-01-01

Nuclear power plant probabilistic risk assessment (PRA) studies utilise many models, simplifications and assumptions. Also subjective judgement is widely applied due to lack of actual data. This results in significant uncertainties. Three general types of uncertainties have been identified: (1) parameter uncertainties, (2) modelling uncertainties, and (3) completeness uncertainties. The significance of some of the modelling assumptions and simplifications cannot be investigated by assignment and propagation of parameter uncertainties. In such cases the impact of different options may (and should) be studied by performing sensitivity analyses, which concentrate on the most critical elements. This paper describes several items suitable for close examination by means of application of sensitivity analysis, when performing a level 1 PRA. Sensitivity analyses are performed with respect to: (1) boundary conditions (success criteria, credit for non-safety systems, degree of detail in modelling of support functions), (2) operator actions, (3) treatment of common cause failures (CCFs). The items of main interest are continuously identified in the course of performing a PRA study, as well as by scrutinising the final results. The practical aspects of sensitivity analysis are illustrated by several applications from a recent PRA study. The critical importance of modelling assumptions is also demonstrated by implementation of some modelling features from another level 1 PRA into the reference model. It is concluded that sensitivity analysis leads to insights important for analysts, reviewers and decision makers. (author)

Risk management on nuclear power plant. Application of probabilistic risk assessment

International Nuclear Information System (INIS)

Kojima, Shigeo

2003-01-01

In U.S.A., nuclear safety regulation is moving to risk-informed regulation (RIR), so necessity of a standard to provide contents of probabilistic risk assessment (PRA) constructing its roots has been discussed for a long time. In 1998, the Committee on Nuclear Risk Management (CNRM) of the American Society of Mechanical Engineers (ASME) began to investigate the standard, of which last edition was published as the Standard for Probabilistic Risk Management for Nuclear Power Plant Applications: RA-S-2002 (PRMA) on April, 2002. As in the Committee, the Nuclear Regulatory Commission (NRC), electric power companies, national institutes, PRA specialists, and so on took parts to carry out many discussions with full energies of participants on risk management in U.S.A., the standard was finished after about four years' efforts. In U.S.A., risk management having already used PRA is successfully practiced, U.S.A. is at a stage with more advancing steps of the risk management than Japan is. Here was described on the standard of PRA and a concrete method of the risk management carried out at nuclear power stations. (G.K.)

Cable Hot Shorts and Circuit Analysis in Fire Risk Assessment

International Nuclear Information System (INIS)

LaChance, Jeffrey; Nowlen, Steven P.; Wyant, Frank

1999-01-01

Under existing methods of probabilistic risk assessment (PRA), the analysis of fire-induced circuit faults has typically been conducted on a simplistic basis. In particular, those hot-short methodologies that have been applied remain controversial in regards to the scope of the assessments, the underlying methods, and the assumptions employed. To address weaknesses in fire PRA methodologies, the USNRC has initiated a fire risk analysis research program that includes a task for improving the tools for performing circuit analysis. The objective of this task is to obtain a better understanding of the mechanisms linking fire-induced cable damage to potentially risk-significant failure modes of power, control, and instrumentation cables. This paper discusses the current status of the circuit analysis task

Insights from Guideline for Performance of Internal Flooding Probabilistic Risk Assessment (IFPRA)

International Nuclear Information System (INIS)

Choi, Sun Yeong; Yang, Joo Eon

2009-01-01

An internal flooding (IF) risk assessment refers to the quantitative probabilistic safety assessment (PSA) treatment of flooding as a result of pipe and tank breaks inside the plants, as well as from other recognized flood sources. The industry consensus standard for Internal Events Probabilistic Risk Assessment (ASME-RA-Sb-2005) includes high-level and supporting technical requirements for developing internal flooding probabilistic risk assessment (IFPRA). This industry standard is endorsed in Regulatory Guide 1.200, Revision 1 as an acceptable approach for addressing the risk contribution from IF events for risk informed applications that require U.S. Nuclear Regulatory commission (NRC) approval. In 2006, EPRI published a draft report for IFPRA that addresses the requirements of the ASME PRA consensus standard and have made efforts to refine and update the final EPRI IFPRA guideline. Westinghouse has performed an IFPRA analysis for several nuclear power plants (NPPs), such as Watts Bar and Fort Calhoun, using the draft EPRI guidelines for development of an IFPRA. Proprietary methodologies have been developed to apply the EPRI guidelines. The objectives of the draft report for IFPRA guideline are to: · Provide guidance for PSA practitioners in the performance of the elements of a PRA associated with internal flooding events consistent with the current state of the art for internal flooding PRA · Provide guidance regarding acceptable approaches that is sufficient to meeting the requirements of the ASME PRA Standard associated with internal flooding · Incorporate lessons learned in the performance of internal flooding PRAs including those identified as pilot applications of earlier drafts of this procedures guide The purpose of this paper is to present a vision for domestic nuclear power plants' IFPRA by comparing the method of the draft EPRI guidelines with the existing IFPRA method for domestic NPPs

Insights from Guideline for Performance of Internal Flooding Probabilistic Risk Assessment (IFPRA)

Energy Technology Data Exchange (ETDEWEB)

Choi, Sun Yeong; Yang, Joo Eon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2009-10-15

An internal flooding (IF) risk assessment refers to the quantitative probabilistic safety assessment (PSA) treatment of flooding as a result of pipe and tank breaks inside the plants, as well as from other recognized flood sources. The industry consensus standard for Internal Events Probabilistic Risk Assessment (ASME-RA-Sb-2005) includes high-level and supporting technical requirements for developing internal flooding probabilistic risk assessment (IFPRA). This industry standard is endorsed in Regulatory Guide 1.200, Revision 1 as an acceptable approach for addressing the risk contribution from IF events for risk informed applications that require U.S. Nuclear Regulatory commission (NRC) approval. In 2006, EPRI published a draft report for IFPRA that addresses the requirements of the ASME PRA consensus standard and have made efforts to refine and update the final EPRI IFPRA guideline. Westinghouse has performed an IFPRA analysis for several nuclear power plants (NPPs), such as Watts Bar and Fort Calhoun, using the draft EPRI guidelines for development of an IFPRA. Proprietary methodologies have been developed to apply the EPRI guidelines. The objectives of the draft report for IFPRA guideline are to: {center_dot} Provide guidance for PSA practitioners in the performance of the elements of a PRA associated with internal flooding events consistent with the current state of the art for internal flooding PRA {center_dot} Provide guidance regarding acceptable approaches that is sufficient to meeting the requirements of the ASME PRA Standard associated with internal flooding {center_dot} Incorporate lessons learned in the performance of internal flooding PRAs including those identified as pilot applications of earlier drafts of this procedures guide The purpose of this paper is to present a vision for domestic nuclear power plants' IFPRA by comparing the method of the draft EPRI guidelines with the existing IFPRA method for domestic NPPs.

Optimization method to branch-and-bound large SBO state spaces under dynamic probabilistic risk assessment via use of LENDIT scales and S2R2 sets

International Nuclear Information System (INIS)

Nielsen, Joseph; Tokuhiro, Akira; Khatry, Jivan; Hiromoto, Robert

2014-01-01

Traditional probabilistic risk assessment (PRA) methods have been developed to evaluate risk associated with complex systems; however, PRA methods lack the capability to evaluate complex dynamic systems. In these systems, time and energy scales associated with transient events may vary as a function of transition times and energies to arrive at a different physical state. Dynamic PRA (DPRA) methods provide a more rigorous analysis of complex dynamic systems. Unfortunately DPRA methods introduce issues associated with combinatorial explosion of states. In order to address this combinatorial complexity, a branch-and-bound optimization technique is applied to the DPRA formalism to control the combinatorial state explosion. In addition, a new characteristic scaling metric (LENDIT – length, energy, number, distribution, information and time) is proposed as linear constraints that are used to guide the branch-and-bound algorithm to limit the number of possible states to be analyzed. The LENDIT characterization is divided into four groups or sets – 'state, system, resource and response' (S2R2) – describing reactor operations (normal and off-normal). In this paper we introduce the branch-and-bound DPRA approach and the application of LENDIT scales and S2R2 sets to a station blackout (SBO) transient. (author)

Classroom Assessment Techniques: A Literature Review

Science.gov (United States)

DiCarlo, Kristen; Cooper, Lori

2014-01-01

Effective classroom assessment techniques are directly linked to course objectives and proposed outcomes. Results within formative and summative assessments have been studied in the online learning environment as educators seek to meet objectives with respect to student success in the non-traditional setting. Online classroom assessment techniques…

Research needs for risk-informed, performance-based regulation

International Nuclear Information System (INIS)

Bailey, J.A.

1997-01-01

Palo Verde Nuclear Generating Station has used PRA-derived risk insights for about 10 years now. The plant originally started applying PRA modeling to an auxiliary feedwater system during the initial licensing phases of the plant, and as a result of that, they were able to work with the NRC and apply some graded quality requirements to that particular system. There was a third redundant auxiliary feedwater pump, and they now can treat that system as partially safety related and partially non-safety related. So it was an advance for Palo Verde at that time to be able to make decisions with a PRA and they began learning how to use those techniques. After completing the IPE it became natural for the plant to make a transition into other areas at the plant to look for areas where the insights gained from PRA could be applied into their decision-making processes. Those that the plant embarked upon initially were areas where they could gain operational risk assessment insights. The author goes on to discuss experiences gained in using these techniques to better assess the safety of operations within the plant. In addition he offers comments on areas which need further development and research to make them more applicable to a plant by plant basis

Using risk based tools in emergency response

International Nuclear Information System (INIS)

Dixon, B.W.; Ferns, K.G.

1987-01-01

Probabilistic Risk Assessment (PRA) techniques are used by the nuclear industry to model the potential response of a reactor subjected to unusual conditions. The knowledge contained in these models can aid in emergency response decision making. This paper presents requirements for a PRA based emergency response support system to date. A brief discussion of published work provides background for a detailed description of recent developments. A rapid deep assessment capability for specific portions of full plant models is presented. The program uses a screening rule base to control search space expansion in a combinational algorithm

Taking the Risk Out of Risk Assessment

Science.gov (United States)

2005-01-01

The ability to understand risks and have the right strategies in place when risky events occur is essential in the workplace. More and more organizations are being confronted with concerns over how to measure their risks or what kind of risks they can take when certain events transpire that could have a negative impact. NASA is one organization that faces these challenges on a daily basis, as effective risk management is critical to the success of its missions especially the Space Shuttle missions. On July 29, 1996, former NASA Administrator Daniel Goldin charged NASA s Office of Safety and Mission Assurance with developing a probabilistic risk assessment (PRA) tool to support decisions on the funding of Space Shuttle upgrades. When issuing the directive, Goldin said, "Since I came to NASA [in 1992], we've spent billions of dollars on Shuttle upgrades without knowing how much they improve safety. I want a tool to help base upgrade decisions on risk." Work on the PRA tool began immediately. The resulting prototype, the Quantitative Risk Assessment System (QRAS) Version 1.0, was jointly developed by NASA s Marshall Space Flight Center, its Office of Safety and Mission Assurance, and researchers at the University of Maryland. QRAS software automatically expands the reliability logic models of systems to evaluate the probability of highly detrimental outcomes occurring in complex systems that are subject to potential accident scenarios. Even in its earliest forms, QRAS was used to begin PRA modeling of the Space Shuttle. In parallel, the development of QRAS continued, with the goal of making it a world-class tool, one that was especially suited to NASA s unique needs. From the beginning, an important conceptual goal in the development of QRAS was for it to help bridge the gap between the professional risk analyst and the design engineer. In the past, only the professional risk analyst could perform, modify, use, and perhaps even adequately understand PRA. NASA wanted

Precision Learning Assessment: An Alternative to Traditional Assessment Techniques.

Science.gov (United States)

Caltagirone, Paul J.; Glover, Christopher E.

1985-01-01

A continuous and curriculum-based assessment method, Precision Learning Assessment (PLA), which integrates precision teaching and norm-referenced techniques, was applied to a math computation curriculum for 214 third graders. The resulting districtwide learning curves defining average annual progress through the computation curriculum provided…

Use of probabilistic risk assessment in maintenance activities at Palo Verde

International Nuclear Information System (INIS)

Lindquist, R.C.; Pobst, D.S.

1993-01-01

Probabilistic risk assessment (PRA) is an important tool in addressing various maintenance activities. At the Palo Verde nuclear generating station (PVNGS), the PRA has been used in a variety of ways to support a wide and diverse selection of maintenance-related activities. For on-line or at-power maintenance, the PRA was used to evaluate combinations of maintenance activities possible with the 12-week or floating maintenance schedule. The maintenance schedule was evaluated to identify any higher risk, undesirable combinations of equipment outages, such as the sole steam-driven auxiliary feedwater pump and the same train emergency diesel generator. Table I is a sampling of the results from the maintenance schedule evaluation in terms of increase in conditional core damage frequency (CDF) above the base- line value due to maintenance on some important key safety systems and combinations thereof. The baseline CDF is 7.4 x 10 -7 per 72 h

Uncertainty analysis in the applications of nuclear probabilistic risk assessment

International Nuclear Information System (INIS)

Le Duy, T.D.

2011-01-01

The aim of this thesis is to propose an approach to model parameter and model uncertainties affecting the results of risk indicators used in the applications of nuclear Probabilistic Risk assessment (PRA). After studying the limitations of the traditional probabilistic approach to represent uncertainty in PRA model, a new approach based on the Dempster-Shafer theory has been proposed. The uncertainty analysis process of the proposed approach consists in five main steps. The first step aims to model input parameter uncertainties by belief and plausibility functions according to the data PRA model. The second step involves the propagation of parameter uncertainties through the risk model to lay out the uncertainties associated with output risk indicators. The model uncertainty is then taken into account in the third step by considering possible alternative risk models. The fourth step is intended firstly to provide decision makers with information needed for decision making under uncertainty (parametric and model) and secondly to identify the input parameters that have significant uncertainty contributions on the result. The final step allows the process to be continued in loop by studying the updating of beliefs functions given new data. The proposed methodology was implemented on a real but simplified application of PRA model. (author)

Probabilistic risk assessment of insecticide concentrations in agricultural surface waters: a critical appraisal.

Science.gov (United States)

Stehle, Sebastian; Knäbel, Anja; Schulz, Ralf

2013-08-01

Due to the specific modes of action and application patterns of agricultural insecticides, the insecticide exposure of agricultural surface waters is characterized by infrequent and short-term insecticide concentration peaks of high ecotoxicological relevance with implications for both monitoring and risk assessment. Here, we apply several fixed-interval strategies and an event-based sampling strategy to two generalized and two realistic insecticide exposure patterns for typical agricultural streams derived from FOCUS exposure modeling using Monte Carlo simulations. Sampling based on regular intervals was found to be inadequate for the detection of transient insecticide concentrations, whereas event-triggered sampling successfully detected all exposure incidences at substantially lower analytical costs. Our study proves that probabilistic risk assessment (PRA) concepts in their present forms are not appropriate for a thorough evaluation of insecticide exposure. Despite claims that the PRA approach uses all available data to assess exposure and enhances risk assessment realism, we demonstrate that this concept is severely biased by the amount of insecticide concentrations below detection limits and therefore by the sampling designs. Moreover, actual insecticide exposure is of almost no relevance for PRA threshold level exceedance frequencies and consequential risk assessment outcomes. Therefore, we propose a concept that features a field-relevant ecological risk analysis of agricultural insecticide surface water exposure. Our study quantifies for the first time the environmental and economic consequences of inappropriate monitoring and risk assessment concepts used for the evaluation of short-term peak surface water pollutants such as insecticides.

Validation of seismic probabilistic risk assessments of nuclear power plants

International Nuclear Information System (INIS)

Ellingwood, B.

1994-01-01

A seismic probabilistic risk assessment (PRA) of a nuclear plant requires identification and information regarding the seismic hazard at the plant site, dominant accident sequences leading to core damage, and structure and equipment fragilities. Uncertainties are associated with each of these ingredients of a PRA. Sources of uncertainty due to seismic hazard and assumptions underlying the component fragility modeling may be significant contributors to uncertainty in estimates of core damage probability. Design and construction errors also may be important in some instances. When these uncertainties are propagated through the PRA, the frequency distribution of core damage probability may span three orders of magnitude or more. This large variability brings into question the credibility of PRA methods and the usefulness of insights to be gained from a PRA. The sensitivity of accident sequence probabilities and high-confidence, low probability of failure (HCLPF) plant fragilities to seismic hazard and fragility modeling assumptions was examined for three nuclear power plants. Mean accident sequence probabilities were found to be relatively insensitive (by a factor of two or less) to: uncertainty in the coefficient of variation (logarithmic standard deviation) describing inherent randomness in component fragility; truncation of lower tail of fragility; uncertainty in random (non-seismic) equipment failures (e.g., diesel generators); correlation between component capacities; and functional form of fragility family. On the other hand, the accident sequence probabilities, expressed in the form of a frequency distribution, are affected significantly by the seismic hazard modeling, including slopes of seismic hazard curves and likelihoods assigned to those curves

Reliability and comparison of acromion assessment techniques on x-ray and magnetic resonance imaging (reliability of acromion assessment techniques)

International Nuclear Information System (INIS)

Viskontas, D.G.; MacDermid, J.C.; Drosdowech, D.S.; Garvin, G.J.; Romano, W.M.; Faber, K.J.

2005-01-01

To determine the reliability and correlation of plain radiography and magnetic resonance imaging (MRI) in the assessment of acromion morphology. Materials and Methods: Acromion morphology was assessed using the lateral acromion angle (LAA) and the acromion-humeral interval (AHI). Thirty patients who had x-rays and MRI for impingement syndrome were included. Six blinded observers assessed the acromion morphology subjectively and objectively. Results: Neither acromion assessment technique demonstrated a positive correlation (kappa and intraclass coefficient 0.55) when measured objectively by experienced observers. Conclusion: The LAA and the AHI are both reliable acromion assessment techniques on X-ray and MRI when measured objectively and by experienced observers. (author)

Uncertainty estimation in nuclear power plant probabilistic safety assessment

International Nuclear Information System (INIS)

Guarro, S.B.; Cummings, G.E.

1989-01-01

Probabilistic Risk Assessment (PRA) was introduced in the nuclear industry and the nuclear regulatory process in 1975 with the publication of the Reactor Safety Study by the U.S. Nuclear Regulatory Commission. Almost fifteen years later, the state-of-the-art in this field has been expanded and sharpened in many areas, and about thirty-five plant-specific PRAs (Probabilistic Risk Assessments) have been performed by the nuclear utility companies or by the U.S. Nuclear Regulatory commission. Among the areas where the most evident progress has been made in PRA and PSA (Probabilistic Safety Assessment, as these studies are more commonly referred to in the international community outside the U.S.) is the development of a consistent framework for the identification of sources of uncertainty and the estimation of their magnitude as it impacts various risk measures. Techniques to propagate uncertainty in reliability data through the risk models and display its effect on the top level risk estimates were developed in the early PRAs. The Seismic Safety Margin Research Program (SSMRP) study was the first major risk study to develop an approach to deal explicitly with uncertainty in risk estimates introduced not only by uncertainty in component reliability data, but by the incomplete state of knowledge of the assessor(s) with regard to basic phenomena that may trigger and drive a severe accident. More recently NUREG-1150, another major study of reactor risk sponsored by the NRC, has expanded risk uncertainty estimation and analysis into the realm of model uncertainty related to the relatively poorly known post-core-melt phenomena which determine the behavior of the molten core and of the rector containment structures

77 FR 61446 - Proposed Revision Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors

Science.gov (United States)

2012-10-09

... Severe Accident Evaluation for New Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Standard... its Standard Review Plan (SRP), Section 19.0, ``Probabilistic Risk Assessment and Severe Accident... assessment (PRA) information and severe accident assessments for new reactors submitted to support design...

Relationship between regulatory issues and probabilistic risk assessments

International Nuclear Information System (INIS)

Ilberg, D.; Papazoglou, I.

1985-01-01

The objective of this study was to obtain some perspective on the characteristics and the relative number of regulatory issues that are PRA related, i.e., can be effectively addressed by plant specific PRA studies. It was also aimed at developing approaches to resolution of regulatory issues as part of plant specific PRAs. Several ongoing NRC programs include a number of safety-related issues which are applicable to operating plants. A number of these issues include aspects that strongly interact with items addressed in PRA studies. The resolution of several generic issues using PRA studies has already started. A review of over 335 issues included in three NRC programs was conducted: Generic Issue Program (GI); Systematic Evaluation Program (SEP); and TMI Action Plan (TMI). The review identifid 240 items related to PRA, 120 of which were judged to have significant effect on core damage frequency. It is believed that these items can be effectively treated in a PRA study that includes internal and external events

Results of the AP600 advanced plant probabilistic risk assessment

International Nuclear Information System (INIS)

Bueter, T.; Sancaktar, S.; Freeland, J.

1997-01-01

The AP600 Probabilistic Risk Assessment (PRA) includes detailed models of the plant systems, including the containment and containment systems that would be used to mitigate the consequences of a severe accident. The AP600 PRA includes a level 1 analysis (core damage frequency), and a level 2 analysis (environmental consequences), an assessment of the plant vulnerability to accidents caused by fire or floods, and a seismic margins analysis. Numerous sensitivities are included in the AP600 PRA including one that assumes no credit for non-safety plant systems. The core damage frequency for the AP600 of 1.7E-07/year is small compared with other PRAs performed in the nuclear industry. The AP600 large release frequency of 1.8E-08/year is also small and shows the ability of the containment systems to prevent a large release should a severe accident occur. Analyses of potential consequences to the environment from a severe accident show that a release would be small, and that containment still provides significant protection 24 hours after an assumed accident. Sensitivity analyses show that plant risk (as measured by core damage frequency and large release frequency) is not sensitive to the reliability of operator actions. 6 refs., 1 fig., 1 tab

Reflections on performance assessment in the Civilian Radioactive Waste Management Program

International Nuclear Information System (INIS)

Fecht, K.R.; Sonnichsen, J.C.; Woods, T.W.

1988-01-01

The national program to develop a high-level waste management system is passing through a period of transition---legislative, political, organizational, and contractual. With the changes that have already occurred and those still under consideration, it is time to examine what ha been done well, in what areas there are now unfulfilled needs in the program, and what improvements might be helpful in the future. The approach and methods with which to assess the performance of the high-level waste management system deserve and enjoy the utmost respect throughout the technical community. Probabilistic risk assessment (PRA) techniques have been conceived, developed and implemented that enable comparative evaluation of various subsystem alternatives, especially geologic media for the repository. This paper argues that the program has need of additional measures of performance important in decision-making processes and suggests additional analytical tools to help assess that performance

Comparison between Canadian probabilistic safety assessment methods formulated by Atomic Energy of Canada limited and probabilistic risk assessment methods

International Nuclear Information System (INIS)

Shapiro, H.S.; Smith, J.E.

1989-01-01

The procedures used by Atomic Energy of Canada Limited (AECL) to perform probabilistic safety assessments (PRAs) differ somewhat from conventionally accepted probabilistic risk assessment (PRA) procedures used elsewhere. In Canada, PSA is used by AECL as an audit tool for an evolving design. The purpose is to assess the safety of the plant in engineering terms. Thus, the PSA procedures are geared toward providing engineering feedback so that necessary changes can be made to the design at an early stage, input can be made to operating procedures, and test and maintenance programs can be optimized in terms of costs. Most PRAs, by contrast, are performed in plants that are already built. Their main purpose is to establish the core melt frequency and the risk to the public due to core melt. Also, any design modification is very expensive. The differences in purpose and timing between PSA and PRA have resulted in differences in methodology and scope. The PSA procedures are used on all plants being designed by AECL

Evaluation of safety issues on newly regulated nuclear power plant by tsunami-level 1 PRA

International Nuclear Information System (INIS)

Tsuji, Yutaro; Miwa, Shuichiro; Mori, Michitsugu

2014-01-01

The tsunami caused by the Great East Japan Earthquake triggered severe accidents involving the units 1 to 4 at the Fukushima Dai-ichi nuclear power station (NPS). In order to re-operate existing nuclear power plants it should be necessary to reduce the core damage frequency on risk by tsunami. In this work, effects of the off-site power supply installation on resuming operation of nuclear power plants were investigated by utilizing the Tsunami-Level 1 Probability Risk Assessment (PRA). Unit 2 of the Onagawa nuclear power station, which resembled units 2 and 3 of Fukushima Dai-ichi, was selected for PRA. First, event-tree was created for the units of the Onagawa nuclear power station with the safety systems such as Emergency Core Cooling System (ECCS), investigating the plant situation at the time of the earthquake and tsunami occurrences. It was assumed that the magnitude of the tsunami was equivalent to the Great East Japan Earthquake. The accident-analytical progression-time was 36 hours, determined from the core-damage occurrence of the unit 3 of Fukushima Dai-ichi nuclear power station. Failure probabilities were calculated by the fault tree, which was created from the elements listed in the event tree. For the calculation, failure rates reported by the NUCIA (NUClear Information Archives) were primarily utilized. Then, obtained failure probabilities were embedded to the event tree. Core damage probabilities were evaluated by calculating success and failure rates for each accidental progression and scenarios. Restoration of the failed equipment and machineries was not considered in the analysis. Installation of the power supply vehicles at the nuclear power plant site reduced the core damage probability from 2.58×10 -6 to 8.56×10 -7 . However, continued addition of the power supply vehicles could not lower the core damage probability further more. In the case of Unit 2 of Onagawa nuclear power station, there could be a limit to lower the core damage

"VEM PRA RUA": THE POLITICAL AND THE POLITICS ON THE WEB

Directory of Open Access Journals (Sweden)

Benedito Fernando Pereira

2014-12-01

Full Text Available Considering the political and the social division of senses, this paper seeks to check how is the process of signification in a society increasingly challenged by electronic discourse and the ways in which political and policy find themselves affected by it. In order do that, we make the analysis of discursive statements “vem pra rua” and “somos a rede social” that were present in banners and posters in street protests in Brazil in 2013, which were organized and delivered, in large part, by virtual means. We observed that such utterances go through a process of appropriation and re-signification, with shifting meanings that run of market logic, go to the social politics and go back to the market logic. We had also observed how the urban environment is affected by the electronic discourse that now constitutes the ways it makes sense.

Probabilistic safety assessment applications and insights

International Nuclear Information System (INIS)

Hitchler, M.J.; Burns, N.L.; Liparulo, N.J.; Mink, F.J.

1987-01-01

The insights gained through a comparison of seven PRA studies (Italian PUN, Sizewell B, Ringhals 2, Millstone 3, Zion 1 and 2, Oconee 3, and Seabrook) included insights regarding the adequacy of the PRA technology utilized in the studies and the potential areas for improvement and insights regarding the adequacy of plant designs and how PRA has been utilized to enhance the design and operation of nuclear power plants. (orig.)

Application of RIA of PRA, AT II and NPY in typing and therapy of EH patients

International Nuclear Information System (INIS)

Yang Yongqing; Wang Xiaozhou; Jiang Qinian

2001-01-01

Objective: To study the typing and AT II receptor inhibitor therapy for essential hypertension (EH) patients. Methods: Plasma RA, AT II and NPY levels were measured by radioimmunoassay (RIA) in 208 Patients with EH and 100 controls; plasma NPY levels were measured in 40 EH patients before and after AT II receptor inhibitor therapy. The mean coefficient of variation for intra and inter batch-assay were less than 10% and 15% respectively. Results: In 208 EH patients plasma PRA levels were increased, normal and decreased in 17.8%, 71.6% and 10.6% respectively, while in 128 EH patients Plasma AT II levels were increased, normal and decreased in 20.3%, 64.1% and 15.6% respectively. In 69 EH Plasma NPY levels were significantly higher than those in 40 control subjects. (17 grade I EH, 137.3 +- 32.6 pg/mL; 28 grade II EH, 148.5 +- 41.1 pg/mL; 24 grade III EH, 162.4 +- 42.7 pg/mL; 40 controls, 118.5 +- 30.5 pg/mL). In 40 EH patients plasma NPY levels were decreased after AT II receptor inhibitor therapy as the blood pressure decreased. Conclusion: Typing of EH patients according to levels of plasma PRA and AT II is useful in guiding treatment. AT II receptor inhibitors are indicated in those patients with increased plasma levels and NPY levels can be used for appraisal of the treatment efficacy

77 FR 66649 - Proposed Revision to Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors

Science.gov (United States)

2012-11-06

... and Severe Accident Evaluation for New Reactors AGENCY: Nuclear Regulatory Commission. ACTION... Assessment and Severe Accident Evaluation for New Reactors.'' The NRC is extending the public comment period... assessment (PRA) information and severe accident assessments for new reactors submitted to support design...

The implications of probabilistic risk assessment for safety policy

International Nuclear Information System (INIS)

Hayns, M.R.

1987-01-01

The use of PRA results in decision making requires a level of understanding on the part of the decision maker which is higher than that obtaining previously. The most important application of PRA lies not in the final results but in the intermediate results which refer to specific systems and operations. Such intermediate results are of great value either at the design stage or later during operation. One of the most 'visible' uses of PRA results is in comparing calculated plant risks with either proposed acceptability criteria, or with other plant, or even natural events. The capability to perform PRA has been established. Only the incorporation of PRA into the licensing process is lacking. The principal conclusions on the implications of PRA for safety policy are as follows: regardless of its state of development, PRA is the only means available for calculating public risk, being able to quantify risk is important in policy related to risk acceptability and to national energy policy. PRAs will be used to establish research and development priorities. Any hazardous plant can be treated using the same methods. More sophisticated methods will be used for solving engineering problems. (author)

Generic safety insights for inspection of boiling water reactors

International Nuclear Information System (INIS)

Higgins, J.C.; Taylor, J.H.; Fresco, A.N.; Hillman, B.M.

1987-01-01

As the number of operating nuclear power plants (NPP) increases, safety inspection has increased in importance. However, precisely what is important, and what is not important? What should one focus inspection efforts on. Over the last two years Probabilistic Risk Assessment (PR) techniques have been developed to aid in the inspection process. Broad interest in generic PRA-based methods has arisen in the past year, since only about 25% of the US nuclear power plants have completed PRAs, and also, inspectors want PRA-based tools for these plants. This paper describes the BNL program to develop generic BWR PRA-based inspection insights or inspection guidance designed to be applied to plants without PRAs

Handbook for value-impact assessments of NRC regulatory actions

International Nuclear Information System (INIS)

Mullen, M.F.; DiPalo, A.J.

1985-01-01

According to current Nuclear Regulatory Commission (NRC) procedures, value-impact (cost-benefit) assessments must be prepared for all rulemaking actions and for a broad range of other regulatory requirements and guidance. Probabilistic risk assessment (PRA) methods furnish an important part of the information base for these assessments. PRA methods are frequently the principal quantitative tool for estimating the benefits (e.g., public risk reduction) of proposed regulatory actions. In December 1983, the NRC published A Handbook for Value-Impact Assessment, NUREG/CR-3568, which provides a set of systematic procedures for performing value-impact assessments. The Handbook contains methods, data, and sources of information that can assist the regulatory analyst in conducting such assessments. The use of probabilistic risk analysis to estimate the benefits of proposed regulatory actions is described. Procedures and methods are also given for evaluating the costs and other consequences associated with regulatory actions. The Handbook has been adopted by the NRC as the recommended guideline for value impact assessments. This paper presents the background, objectives, and scope of the Handbook, describes the value-impact assessment methods (including the use of probabilistic risk assessment to estimate benefits), and discusses a selection of current and planned applications, with examples to illustrate how the methods are used

Probabilistic Risk Assessment Procedures Guide for NASA Managers and Practitioners (Second Edition)

Science.gov (United States)

Stamatelatos,Michael; Dezfuli, Homayoon; Apostolakis, George; Everline, Chester; Guarro, Sergio; Mathias, Donovan; Mosleh, Ali; Paulos, Todd; Riha, David; Smith, Curtis;

Advanced Test Reactor probabilistic risk assessment methodology and results summary

International Nuclear Information System (INIS)

Eide, S.A.; Atkinson, S.A.; Thatcher, T.A.

1992-01-01

The Advanced Test Reactor (ATR) probabilistic risk assessment (PRA) Level 1 report documents a comprehensive and state-of-the-art study to establish and reduce the risk associated with operation of the ATR, expressed as a mean frequency of fuel damage. The ATR Level 1 PRA effort is unique and outstanding because of its consistent and state-of-the-art treatment of all facets of the risk study, its comprehensive and cost-effective risk reduction effort while the risk baseline was being established, and its thorough and comprehensive documentation. The PRA includes many improvements to the state-of-the-art, including the following: establishment of a comprehensive generic data base for component failures, treatment of initiating event frequencies given significant plant improvements in recent years, performance of efficient identification and screening of fire and flood events using code-assisted vital area analysis, identification and treatment of significant seismic-fire-flood-wind interactions, and modeling of large loss-of-coolant accidents (LOCAs) and experiment loop ruptures leading to direct damage of the ATR core. 18 refs

As praças dr. Augusto Silva e Leonardo Venerando Pereira, 1701 Lavras - MG, segundo a visão dos seus freqüentadores The park dr. Augusto Silva, Lavras - MG, according to viewpoint of its users

Directory of Open Access Journals (Sweden)

Alessandra Teixeira da Silva

2008-12-01

Full Text Available As praças públicas, além de contribuírem para o embelezamento das cidades, também desempenham uma função importante no contexto urbanístico ambiental. A Praça Dr. Augusto Silva, situada no município de Lavras, MG, também já foi chamada de Largo da Matriz, Praça Central e Jardim Municipal, que foi inaugurado oficialmente em 29 de novembro de 1908, quando passou a ter o nome de um ilustre médico lavrense. A praça Dr. Augusto Silva, possui atualmente, área de 7.552,65 m² e apresenta como prolongamento a praça Leonardo Venerando Pereira, com área atual de 2.041,72 m². Até 1940, esse prolongamento era denominado de Praça da Bandeira. Constatou-se, que desde o início da década de 1910, o local foi palco de grandes celebrações e encontros políticos e também foi muito freqüentado pela população. A Praça possui rica vegetação, onde se destacam uma centenária Tipuana (Tipuana tipu, diversos ipês (Tabebuia spp., palmeiras-imperiais (Roystonea oleracea. Objetivando-se avaliar a situação atual e usos, foi realizada uma avaliação, mediante análise quantitativa e pesquisa junto à população. Foram feitas uma série de avaliações referentes aos seguintes itens: aspectos urbanísticos, físicos e a vegetação. A pesquisa de opinião foi aplicada a aproximadamente 600 usuários, na própria praça, em dias da semana e horários diferentes, por meio de um questionário com perguntas diretas ao próprio entrevistado. Os dados coletados foram analisados estatisticamente utilizando o Software SPSS, de onde foram obtidas as freqüências porcentuais. A maior parte dos entrevistados julga a praça como um ponto de encontro entre amigos, onde contemplam a sua beleza e descansam. Por meio de pesquisa de opinião realizada com os usuários da praça, pôde-se identificar que se trata de um local muito freqüentado por várias faixas etárias, em diferentes períodos do dia. A praça Dr. Augusto Silva mostra-se de grande import

Combining objective and subjective techniques for assessing quality of management

International Nuclear Information System (INIS)

Arueti, S.; Okrent, D.

1987-01-01

The basic assumption is that utility and plant corporate management have a significant role in plant safety which may be quantifiable. From this point of view we try to identify symptoms and paths through which management effectiveness affects plant safety, partly in terms of measureable parameters. Some of the available data are analyzed in light of the proposed parameters, in order to examine possible correlations. Preliminary proposals are made of methods for including management performance as a variable in future PRA studies. This paper focuses primarily on measures of management quality from relating to what are frequently called performance indicators, such as SALP ratings, number of scrams, ESF actuations and safety system failures and challanges; forced outages; availability; enforcement actions; and licensee event reports (LERs). (orig./HP)

Use of probabilistic risk assessment in expert system usage for nuclear power plant safety

International Nuclear Information System (INIS)

Uhrig, R.E.

1987-01-01

The introduction of probability risk assessments (PRA's) to nuclear power plants in the Rasmussen Report (WASH-1400) gave us a means of evaluating the risk to the public associated with the operation of nuclear power plants, at least on a relative basis. While the choice of the ''source term'' and methodology in a PRA significantly influence the absolute probability and the consequences of core melt, comparison of two PRA calculations for two configurations of the same plant, carried out on a consistent basis, can be readily identify the increase in risk associated with going from one configuration of a plant to another by removing components or systems from service. This ratio of core melt probabilities (assuming no recovery of failed systems) obtained from two PRA calculations for different configurations was the criterion (called ''risk factor'') chosen as a basis for making a decision in an expert system as to what mitigating action, if any, would be taken to avoid a trip situation from developing. PRISIM was developed by JBF Associates of Knoxville under the sponsorship of the NRC as a system for Resident Inspectors at nuclear power plants to provide them with a relative safety status of the plant under all configurations. PRISIM calculated the risk factor---the ration of core melt probabilities of the plant under the current configuration relative to the normal configuration with all systems functioning---using an algorithm that emulates the results of the original PRA. It also presents time and core melt (assuming no recovery of systems or components)

Probabilistic safety assessment support for the maintenance rule at Duke Power Company

International Nuclear Information System (INIS)

Brewer, H. Duncan; Canady, Ken S.

1999-01-01

The Nuclear Regulatory Commission (NRC) published the Maintenance Rule on July 10, 1991 with an implementation date of July 10, 1996 . Maintenance rule implementation at the Duke Power Company has used probabilistic safety assessment (PSA) insights to help focus the monitoring of structures, systems and components (SSC) performance and to ensure that maintenance is effectively performed. This paper describes how the probabilistic risk assessment (PRA) group at the Duke Power Company provides support for the maintenance rule by performing the following tasks: (1) providing a member of the expert panel; (2) determining the risk-significant SSCs; (3) establishing SSC performance criteria for availability and reliability; (4) evaluating past performance and its impact on core damage risk as part of the periodic assessment; (5) providing input to the PRA matrix; (6) providing risk analyses of combinations of SSCs out of service; (7) providing support for the SENTINEL program; and (8) providing support for PSA training. These tasks are not simply tied to the initial implementation of the rule. The maintenance rule must be kept consistent with the current design and operation of the plant. This will require that the PRA models and the many PSA calculations performed to support the maintenance rule are kept up-to-date. Therefore, support of the maintenance rule will be one of the primary roles of the PSA group for the remainder of the life of the plant

Assessment Techniques in the Second-Cycle Studies

Directory of Open Access Journals (Sweden)

Jurgita Kerevičienė

2011-06-01

Full Text Available Assessment and evaluation are among the essential functions performed by an educator. What is more, they are closely related to the communication of the teaching process efficiency in terms of verification of learner and educator expectations, determination of learning and teaching quality achievement, and generalization of educational activities. The present study aims at the analysis of assessment techniques applied in the master degree linguistic studies. Initially, a brief theoretical overview of both formal and informal assessment me- thods is provided, with the focus on their distinctive features and ways of application in the light of Bloom’s taxonomy and Dublin description representing the competence requirements set for the gain of definite skills. Further, the results of the survey (conducted at Kaunas Faculty of Humanities, Vilnius University are described. Finally, the eligible assessment techniques serving the purpose of efficient educational process in the second-cycle study process are identified.

Probablistic risk assessment methodology application to Indian pressurised heavy water reactors

International Nuclear Information System (INIS)

Babar, A.K.; Grover, R.B.; Mehra, V.K.; Gangwal, D.K.; Chakraborty, G.

1987-01-01

Probabilistic risk assessment in the context of nuclear power plants is associated with models that predict the offsite radiological releases resulting from reactor accidents. Level 1 PRA deals with the identification of accident sequences relevant to the design of a system and also with their quantitative estimation. It is characterised by event tree, fault tree analysis. The initiating events applicable to pressurised heavy water reactors have been considered and the dominating initiating events essential for detailed studies are identified in this paper. Reliability analysis and the associated problems encountered during the case studies are mentioned briefly. It is imperative to validate the failure data used for analysis. Bayesian technique has been employed for the same and a brief account is included herein. A few important observations, e.g. effects of the presence of moderator, made during the application of probabilistic risk assessment methodology are also discussed. (author)

Human Reliability Analysis in Support of Risk Assessment for Positive Train Control

Science.gov (United States)

2003-06-01

This report describes an approach to evaluating the reliability of human actions that are modeled in a probabilistic risk assessment : (PRA) of train control operations. This approach to human reliability analysis (HRA) has been applied in the case o...

Risk assessment application to NRC inspection

International Nuclear Information System (INIS)

Campbell, D.J.; Guthrie, V.H.; Flanagan, G.F.

1987-01-01

Inspectors must make many decisions on the allocation of their efforts. To date, these decisions have been made based upon their own judgment and guidance from inspection procedures. The program described in this paper provides PRA information as an additional aid to inspectors. A structured approach for relating PRA information to specific inspection decisions has been developed. The use of PRA information as an aid in optimal decision making (1) in response to the current plant status and (2) in the scheduling of effort over an extended period of time is considered. (orig.)

Technical Needs for Enhancing Risk Monitors with Equipment Condition Assessment for Advanced Small Modular Reactors

Energy Technology Data Exchange (ETDEWEB)

Coble, Jamie B.; Coles, Garill A.; Ramuhalli, Pradeep; Meyer, Ryan M.; Berglin, Eric J.; Wootan, David W.; Mitchell, Mark R.

2013-04-04

Advanced small modular reactors (aSMRs) can provide the United States with a safe, sustainable, and carbon-neutral energy source. The controllable day-to-day costs of aSMRs are expected to be dominated by operation and maintenance costs. Health and condition assessment coupled with online risk monitors can potentially enhance affordability of aSMRs through optimized operational planning and maintenance scheduling. Currently deployed risk monitors are an extension of probabilistic risk assessment (PRA). For complex engineered systems like nuclear power plants, PRA systematically combines event likelihoods and the probability of failure (POF) of key components, so that when combined with the magnitude of possible adverse consequences to determine risk. Traditional PRA uses population-based POF information to estimate the average plant risk over time. Currently, most nuclear power plants have a PRA that reflects the as-operated, as-modified plant; this model is updated periodically, typically once a year. Risk monitors expand on living PRA by incorporating changes in the day-by-day plant operation and configuration (e.g., changes in equipment availability, operating regime, environmental conditions). However, population-based POF (or population- and time-based POF) is still used to populate fault trees. Health monitoring techniques can be used to establish condition indicators and monitoring capabilities that indicate the component-specific POF at a desired point in time (or over a desired period), which can then be incorporated in the risk monitor to provide a more accurate estimate of the plant risk in different configurations. This is particularly important for active systems, structures, and components (SSCs) proposed for use in aSMR designs. These SSCs may differ significantly from those used in the operating fleet of light-water reactors (or even in LWR-based SMR designs). Additionally, the operating characteristics of aSMRs can present significantly different

Analysis of climate and anthropogenic impacts on runoff in the Lower Pra River Basin of Ghana.

Science.gov (United States)

Awotwi, Alfred; Anornu, Geophrey Kwame; Quaye-Ballard, Jonathan; Annor, Thompson; Forkuo, Eric Kwabena

2017-12-01

The Lower Pra River Basin (LPRB), located in the forest zone of southern Ghana has experienced changes due to variability in precipitation and diverse anthropogenic activities. Therefore, to maintain the functions of the ecosystem for water resources management, planning and sustainable development, it is important to differentiate the impacts of precipitation variability and anthropogenic activities on stream flow changes. We investigated the variability in runoff and quantified the contributions of precipitation and anthropogenic activities on runoff at the LPRB. Analysis of the precipitation-runoff for the period 1970-2010 revealed breakpoints in 1986, 2000, 2004 and 2010 in the LPRB. The periods influenced by anthropogenic activities were categorized into three periods 1987-2000, 2001-2004 and 2005-2010, revealing a decrease in runoff during 1987-2000 and an increase in runoff during 2001-2004 and 2005-2010. Assessment of monthly, seasonal and annual runoff depicted a significant increasing trend in the runoff time series during the dry season. Generally, runoff increased at a rate of 9.98 × 10 7 m 3 yr -1 , with precipitation variability and human activities contributing 17.4% and 82.3% respectively. The dominant small scale alluvial gold mining activity significantly contributes to the net runoff variability in LPRB.

Pra que time ele joga?: a produção da identidade homossexual em um vídeo educativo

Directory of Open Access Journals (Sweden)

Nilson Fernandes Dinis

2009-01-01

Full Text Available This work aims to analyze the discourse about homosexuality by means of a research with undergraduate students of Physical Education. For this purpose, it made use of the film "Pra que time ele joga?" as a tool of discussing the theme with the research group, aiming to observe what kind of opinions these students have about homosexuality, as well as the discursive production of the image of homosexual subject produced by the educational videotape.

Integrated fault tree development environment

International Nuclear Information System (INIS)

Dixon, B.W.

1986-01-01

Probabilistic Risk Assessment (PRA) techniques are utilized in the nuclear industry to perform safety analyses of complex defense-in-depth systems. A major effort in PRA development is fault tree construction. The Integrated Fault Tree Environment (IFTREE) is an interactive, graphics-based tool for fault tree design. IFTREE provides integrated building, editing, and analysis features on a personal workstation. The design philosophy of IFTREE is presented, and the interface is described. IFTREE utilizes a unique rule-based solution algorithm founded in artificial intelligence (AI) techniques. The impact of the AI approach on the program design is stressed. IFTREE has been developed to handle the design and maintenance of full-size living PRAs and is currently in use

Engineering aspects of probabilistic risk assessment

International Nuclear Information System (INIS)

vonHerrmann, J.L.; Wood, P.J.

1984-01-01

Over the last decade, the use of probabilistic risk assessment (PRA) in the nuclear industry has expanded significantly. In these analyses the probabilities of experiencing certain undesired events (for example, a plant accident which results in damage to the nuclear fuel) are estimated and the consequences of these events are evaluated in terms of some common measure. These probabilities and consequences are then combined to form a representation of the risk associated with the plant studied. In the relatively short history of probabilistic risk assessment of nuclear power plants, the primary motivation for these studies has been the quantitative assessment of public risk associated with a single plant or group of plants. Accordingly, the primary product of most PRAs performed to date has been a 'risk curve' in which the probability (or expected frequency) of exceeding a certain consequence level is plotted against that consequence. The most common goal of these assessments has been to demonstrate the 'acceptability' of the calculated risk by comparison of the resultant risk curve to risk curves associated with other plants or with other societal risks. Presented here are brief descriptions of some alternate applications of PRAs, a discussion of how these other applications compare or contrast with the currently popular uses of PRA, and a discussion of the relative benefits of each

Probabilistic risk assessment: Number 219

International Nuclear Information System (INIS)

Bari, R.A.

1985-01-01

This report describes a methodology for analyzing the safety of nuclear power plants. A historical overview of plants in the US is provided, and past, present, and future nuclear safety and risk assessment are discussed. A primer on nuclear power plants is provided with a discussion of pressurized water reactors (PWR) and boiling water reactors (BWR) and their operation and containment. Probabilistic Risk Assessment (PRA), utilizing both event-tree and fault-tree analysis, is discussed as a tool in reactor safety, decision making, and communications. (FI)

System Analysis and Risk Assessment System (SARA), Version 4.0

International Nuclear Information System (INIS)

Russell, K.D.; Sattison, M.B.; Skinner, N.L.; Stewart, H.D.; Wood, S.T.

1992-02-01

This NUREG is the reference manual for the System Analysis and Risk Assessment (SARA) System Version 4.0, a microcomputer-based system used to analyze the safety issues of a family [i.e., a power plant, a manufacturing facility, any facility on which a probabilistic risk assessment (PRA) might be performed]. The SARA data base contains PRA data for the dominant accident sequences of a family and descriptive information about the family including event trees, fault trees, and system model diagrams. The number of facility data bases that can be accessed is limited only by the amount of disk storage available. To simulate changes to family systems, SARA users change the failure rates of initiating and basic events and/or modify the structure of the cut sets that make up the event trees, fault trees, and systems. The user then evaluates the effects of these changes through the recalculation of the resultant accident sequence probabilities and importance measures. The results are displayed in tables and graphs

The capabilities and applications of the saphire 5.0 safety assessment software

International Nuclear Information System (INIS)

Russell, K.D.; Wood, S.T.; Kvarfordt, K.J.

1994-01-01

The System Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) refers to a suite of computer programs that were developed to create and analyze a probabilistic risk assessment (PRA) of a nuclear power plant. The programs in this suite include: Models and Results Data Base (MAR-D) software, Integrated Reliability and Risk Analysis System (IRRAS) software, System Analysis and Risk Assessment (SARA) software, and Fault tree, Event tree, and Piping and instrumentation diagram (FEP) graphical editor. Each of these programs performs a specific function in taking a PRA from the conceptual state all the way to publication. This paper provides an overview of the features and capabilities provided in version 5.0 of this software system. Some major new features include the ability to store unlimited cut sets, the ability to perform location transformations, the ability to perform seismic analysis, the ability to perform automated rule based recovery analysis and end state cut set partitioning, the ability to perform end state analysis, a new alphanumeric fault tree editor, and a new alphanumeric event tree editor. Many enhancements and improvements to the user interface as well as a significant reduction in the time required to perform an analysis are included in version 5.0. These new features and capabilities provide a powerful set of PC based PRA analysis tools

Effectiveness of teachers' training in assessment techniques: participants' perception.

Science.gov (United States)

Baral, N; Gautam, A; Lamsal, M; Paudel, B H; Lal Das, B K; Aryal, M

2011-01-01

Medical education has experienced a rapidly changing scenario especially during the past three decades. Teachers training programs have emerged in recent years as a more in depth and comprehensive mechanisms for strengthening instructional skills and formulating assessment techniques among faculty members. The present study was carried out with the aim of assessing effectiveness of teacher training workshop on assessment techniques at B.P. Koirala Institute of Health sciences (BPKIHS), Dharan, Nepal. Methods: In this prospective questionnaire based study, the 26 entry level teachers were enrolled. They were requested to fill the questionnaire before and after the workshop on the assessment techniques at BPKIHS, Dharan. Twenty six teachers majority of whom were entry level teachers comprising of medical, dental, nursing and allied sciences participated in pre- and post test analysis on the various components of written and oral examinations. A two and half days long training was conducted having 9 hours of teaching interactive sessions, 6 hours of group exercises, daily evaluation and 3 hours of presentation by participants. There was high degree of agreement with most topics of the assessment training. There was a significant gain in knowledge (p training being informative and learned new things about assessment. More practice session with longer duration with invited resource persons were also suggested for an effective session. Thereis an immense opportunity for a better and comprehensive training in evaluation techniques in medical education..

Advanced Test Reactor outage risk assessment

International Nuclear Information System (INIS)

Thatcher, T.A.; Atkinson, S.A.

1997-01-01

Beginning in 1997, risk assessment was performed for each Advanced Test Reactor (ATR) outage aiding the coordination of plant configuration and work activities (maintenance, construction projects, etc.) to minimize the risk of reactor fuel damage and to improve defense-in-depth. The risk assessment activities move beyond simply meeting Technical Safety Requirements to increase the awareness of risk sensitive configurations, to focus increased attention on the higher risk activities, and to seek cost-effective design or operational changes that reduce risk. A detailed probabilistic risk assessment (PRA) had been performed to assess the risk of fuel damage during shutdown operations including heavy load handling. This resulted in several design changes to improve safety; however, evaluation of individual outages had not been performed previously and many risk insights were not being utilized in outage planning. The shutdown PRA provided the necessary framework for assessing relative and absolute risk levels and assessing defense-in-depth. Guidelines were written identifying combinations of equipment outages to avoid. Screening criteria were developed for the selection of work activities to receive review. Tabulation of inherent and work-related initiating events and their relative risk level versus plant mode has aided identification of the risk level the scheduled work involves. Preoutage reviews are conducted and post-outage risk assessment is documented to summarize the positive and negative aspects of the outage with regard to risk. The risk for the outage is compared to the risk level that would result from optimal scheduling of the work to be performed and to baseline or average past performance

A pilot with computer-assisted psychosocial risk –assessment for refugees

Directory of Open Access Journals (Sweden)

Ahmad Farah

2012-07-01

Full Text Available Abstract Background Refugees experience multiple health and social needs. This requires an integrated approach to care in the countries of resettlement, including Canada. Perhaps, interactive eHealth tools could build bridges between medical and social care in a timely manner. The authors developed and piloted a multi-risk Computer-assisted Psychosocial Risk Assessment (CaPRA tool for Afghan refugees visiting a community health center. The iPad based CaPRA survey was completed by the patients in their own language before seeing the medical practitioner. The computer then generated individualized feedback for the patient and provider with suggestions about available services. Methods A pilot randomized trial was conducted with adult Afghan refugees who could read Dari/Farsi or English language. Consenting patients were randomly assigned to the CaPRA (intervention or usual care (control group. All patients completed a paper-pencil exit survey. The primary outcome was patient intention to see a psychosocial counselor. The secondary outcomes were patient acceptance of the tool and visit satisfaction. Results Out of 199 approached patients, 64 were eligible and 50 consented and one withdrew (CaPRA = 25; usual care = 24. On average, participants were 37.6 years of age and had lived 3.4 years in Canada. Seventy-two percent of participants in CaPRA group had intention to visit a psychosocial counselor, compared to 46 % in usual care group [X2 (1=3.47, p = 0.06]. On a 5-point scale, CaPRA group participants agreed with the benefits of the tool (mean = 4 and were ‘unsure’ about possible barriers to interact with the clinicians (mean = 2.8 or to privacy of information (mean = 2.8 in CaPRA mediated visits. On a 5-point scale, the two groups were alike in patient satisfaction (mean = 4.3. Conclusion The studied eHealth tool offers a promising model to integrate medical and social care to address the health and settlement

Devonian conodonts from the Foča–Prača Paleozoic complex (Durmitor Nappe, southeastern Bosnia and Herzegovina

Directory of Open Access Journals (Sweden)

Bogdan Jurkovšek

2011-06-01

Full Text Available Conodont study of the Crna Rijeka borehole CR-17, positioned in the frontal part of the Durmitor Nappe (Foča – Prača Paleozoic complex, SE Bosnia and Herzegovina is presented. The obtained fauna indicates an Early-Middle Devonian age and due to poor preservation an identification at a generic level is possible only. The recovered conodont elements have a high Color Alteration Index (CAI = 6,5–7 indicating a degree of metamorphism corresponding to a temperature interval from 440 °C to 720 °C.

Survey of probabilistic methods in safety and risk assessment for nuclear power plant licensing

International Nuclear Information System (INIS)

1984-04-01

After an overview about the goals and general methods of probabilistic approaches in nuclear safety the main features of probabilistic safety or risk assessment (PRA) methods are discussed. Mostly in practical applications not a full-fledged PRA is applied but rather various levels of analysis leading from unavailability assessment of systems over the more complex analysis of the probable core damage stages up to the assessment of the overall health effects on the total population from a certain practice. The various types of application are discussed in relation to their limitation and benefits for different stages of design or operation of nuclear power plants. This gives guidance for licensing staff to judge the usefulness of the various methods for their licensing decisions. Examples of the application of probabilistic methods in several countries are given. Two appendices on reliability analysis and on containment and consequence analysis provide some more details on these subjects. (author)

Ovos de Toxocara sp. e larvas de Ancylostoma sp. em praça pública de Lavras, MG Toxocara sp. eggs and Ancylostoma sp. larva in public parks, Brazil

Directory of Open Access Journals (Sweden)

Antônio Marcos Guimarães

2005-04-01

Full Text Available Larva migrans visceral e cutânea são zoonoses parasitárias causadas pela infecção da larva de Toxocara sp. e Ancylostoma sp., respectivamente. O objetivo do estudo foi verificar a contaminação por ovos de Toxocara sp. e ovos e larvas de Ancylostoma sp. em amostras de solos coletadas de praças públicas e de áreas de recreação infantil de Lavras, Estado de Minas Gerais, por meio da técnica de centrífugo-flutuação e do método de Baermann. A ocorrência de ovos de Toxocara sp. e, ovos e larvas de Ancylostoma sp. foi observada em 69,6% (16/23 das amostras de solo coletadas de praças públicas. A contaminação somente por ovos de Ancylostoma sp. em amostras de solo coletadas em escolas/creches foi de 22,2% (4/18. A percentagem de amostras de areia coletadas de escolas/creches contaminadas somente com larvas de Ancylostoma sp. foi de 11,1% (2/18. Praças públicas são as áreas com maior risco potencial de infecção por Toxocara sp. e Ancylostoma sp. Exame coproparasitológico realizado em 174 amostras de fezes de cães observou 58% e 23%, respectivamente, com ovos de Ancylostoma sp. e Toxocara sp.Visceral and cutaneous larva migrans are parasitic zoonoses caused by the infection of larval nematodes Toxocara sp. and Ancylostoma sp. respectively. The objective of this study was to investigate the contamination by Toxocara sp. eggs and Ancylostoma sp. eggs and larva of soil samples collected from public parks and children's playground areas in state of Minas Gerais, Brazil, using both Baermann's method and centrifugal flotation technique. Toxocara sp. and Ancylostoma sp. eggs were observed in soil samples collected from public squares in 17.4% (4/23 and 69.6 (16/23 respectively. In schools and child day care settings the contamination by Ancylostoma sp. larva in sand samples was 11.1% (2/18. Public parks are settings of more potential risk of Toxocara sp. eggs and Ancylostoma sp. infection. Stool parasitology testing of 174 stool

Objective techniques for psychological assessment

Science.gov (United States)

Wortz, E.; Hendrickson, W.; Ross, T.

1973-01-01

A literature review and a pilot study are used to develop psychological assessment techniques for determining objectively the major aspects of the psychological state of an astronaut. Relationships between various performance and psychophysiological variables and between those aspects of attention necessary to engage successfully in various functions are considered in developing a paradigm to be used for collecting data in manned isolation chamber experiments.

Multi-intelligence critical rating assessment of fusion techniques (MiCRAFT)

Science.gov (United States)

Blasch, Erik

2015-06-01

Assessment of multi-intelligence fusion techniques includes credibility of algorithm performance, quality of results against mission needs, and usability in a work-domain context. Situation awareness (SAW) brings together low-level information fusion (tracking and identification), high-level information fusion (threat and scenario-based assessment), and information fusion level 5 user refinement (physical, cognitive, and information tasks). To measure SAW, we discuss the SAGAT (Situational Awareness Global Assessment Technique) technique for a multi-intelligence fusion (MIF) system assessment that focuses on the advantages of MIF against single intelligence sources. Building on the NASA TLX (Task Load Index), SAGAT probes, SART (Situational Awareness Rating Technique) questionnaires, and CDM (Critical Decision Method) decision points; we highlight these tools for use in a Multi-Intelligence Critical Rating Assessment of Fusion Techniques (MiCRAFT). The focus is to measure user refinement of a situation over the information fusion quality of service (QoS) metrics: timeliness, accuracy, confidence, workload (cost), and attention (throughput). A key component of any user analysis includes correlation, association, and summarization of data; so we also seek measures of product quality and QuEST of information. Building a notion of product quality from multi-intelligence tools is typically subjective which needs to be aligned with objective machine metrics.

Analysis of dependent failures in risk assessment and reliability evaluation

International Nuclear Information System (INIS)

Fleming, K.N.; Mosleh, A.; Kelley, A.P. Jr.; Gas-Cooled Reactors Associates, La Jolla, CA)

1983-01-01

The ability to estimate the risk of potential reactor accidents is largely determined by the ability to analyze statistically dependent multiple failures. The importance of dependent failures has been indicated in recent probabilistic risk assessment (PRA) studies as well as in reports of reactor operating experiences. This article highlights the importance of several different types of dependent failures from the perspective of the risk and reliability analyst and provides references to the methods and data available for their analysis. In addition to describing the current state of the art, some recent advances, pitfalls, misconceptions, and limitations of some approaches to dependent failure analysis are addressed. A summary is included of the discourse on this subject, which is presented in the Institute of Electrical and Electronics Engineers/American Nuclear Society PRA Procedures Guide

Controlling principles for prior probability assignments in nuclear risk assessment

International Nuclear Information System (INIS)

Cook, I.; Unwin, S.D.

1986-01-01

As performed conventionally, nuclear probabilistic risk assessment (PRA) may be criticized as utilizing inscrutable and unjustifiably ''precise'' quantitative informed judgment or extrapolation from that judgment. To meet this criticism, controlling principles that govern the formulation of probability densities are proposed, given only the informed input that would be required for a simple bounding analysis. These principles are founded upon information theoretic ideas of maximum uncertainty and cover both cases in which there exists a stochastic model of the phenomenon of interest and cases in which these is no such model. In part, the principles are conventional, and such an approach is justified by appealing to certain analogies in accounting practice and judicial decision making. Examples are given. Appropriate employment of these principles is expected to facilitate substantial progress toward PRA scrutability and transparency

Overview of seismic margin insights gained from seismic PRA results

International Nuclear Information System (INIS)

Kennedy, R.P.; Sues, R.H.; Campbell, R.D.

1986-01-01

This paper presents the findings of a study conducted under NRC and EPRI sponsorship in which published seismic PRAs were reviewed in order to gain insight to the seismic margins inherent in existing nuclear plants. The approach taken was to examine the fragilities of those components which have been found to be dominant contributors to seismic risk at plants in low-to-moderate seismic regions (SSE levels between 0.12g and 0.25g). It is concluded that there is significant margin inherent in the capacity of most critical components above the plant design basis. For ground motions less than about 0.3g, the predominant sources of seismic risk are loss of offsite power coupled with random failure of the emergency diesels, non-recoverable circuit breaker trip due to relay chatter, unanchored equipment, unreinforced non-load bearing block walls, vertical water storage tanks, systems interactions and possibly soil liquefaction. Recommendations as to which components should be reviewed in seismic margin studies for margin earthquakes less than 0.3g, between 0.3g and 0.5g, and greater than 0.5g, developed by the NRC expert panel on the quantification of seismic margins (based on the review of past PRA data, earthquake experience data, and their own personal experience) are presented

Cognitive environment simulation: An artificial intelligence system for human performance assessment: Modeling human intention formation: [Technical report, May 1986-June 1987

International Nuclear Information System (INIS)

Woods, D.D.; Roth, E.M.; Pople, H. Jr.

1987-11-01

This report documents the results of Phase II of a three phase research program to develop and validate improved methods to model the cognitive behavior of nuclear power plant (NPP) personnel. In Phase II a dynamic simulation capability for modeling how people form intentions to act in NPP emergency situations was developed based on techniques from artificial intelligence. This modeling tool, Cognitive Environment Simulation or CES, simulates the cognitive processes that determine situation assessment and intention formation. It can be used to investigate analytically what situations and factors lead to intention failures, what actions follow from intention failures (e.g., errors of omission, errors of commission, common mode errors), the ability to recover from errors or additional machine failures, and the effects of changes in the NPP person-machine system. The Cognitive Reliability Assessment Technique (or CREATE) was also developed in Phase II to specify how CES can be used to enhance the measurement of the human contribution to risk in probabilistic risk assessment (PRA) studies. 43 refs., 20 figs., 1 tab

Simplified probabilistic risk assessment in fuel reprocessing

International Nuclear Information System (INIS)

Solbrig, C.W.

1993-01-01

An evaluation was made to determine if a backup mass tracking computer would significantly reduce the probability of criticality in the fuel reprocessing of the Integral Fast Reactor. Often tradeoff studies, such as this, must be made that would greatly benefit from a Probably Risk Assessment (PRA). The major benefits of a complete PRA can often be accrued with a Simplified Probabilistic Risk Assessment (SPRA). An SPRA was performed by selecting a representative fuel reprocessing operation (moving a piece of fuel) for analysis. It showed that the benefit of adding parallel computers was small compared to the benefit which could be obtained by adding parallelism to two computer input steps and two of the weighing operations. The probability of an incorrect material moves with the basic process is estimated to be 4 out of 100 moves. The actual values of the probability numbers are considered accurate to within an order of magnitude. The most useful result of developing the fault trees accrue from the ability to determine where significant improvements in the process can be made. By including the above mentioned parallelism, the error move rate can be reduced to 1 out of 1000

The Sanskrit and Arabic Sources of the Praśnatantra Attributed to Nīlakaṇṭha

Directory of Open Access Journals (Sweden)

Martin Gansten

2014-12-01

Full Text Available The highly popular Praśnatantra attributed to Nīlakaṇṭha of Kāśī (fl. late 16th century and sometimes regarded as the third volume of his Tājikanīlakaṇṭhī is shown to depend for its basic structure on an abridged Sanskrit version of the Kitāb fi l-masāʾil wa-l-aḥkām by Sahl ibn Bishr (early 9th century, apparently authored by Samarasiṃha in the 13th century, to which quotations primarily from Sanskrit astrological works in the classical Indian style have been added, resulting in a hybrid of Indian and Perso-Arabic interrogational astrology.

Review of UCN 3,4 PSA model based on NEI PRA peer review process guidance, rev.0

International Nuclear Information System (INIS)

Yang, Joon Eon; Kang, D. I.; Kim, K. Y.; Lee, Y. H.; Jang, S. C.; Ha, J. J.; Han, S. H.; Han, S. J.; Hwang, M. J.

2003-05-01

Recently, under the de-regulation environment, nuclear industry has attempted various approaches to improve the economics of Nuclear Power Plants (NPP). One of these efforts is the Risk Informed/Performance-Based Operation (RIPBO). This approach uses the risk and performance information to manage the resources effectively and efficiently that are used in the operation of NPP. In RIPBO, PSA quality is one of the most important things. The nuclear industry and regulatory body of U.S.A have developed a measure to evaluate the quality of PSA. NEI (Nuclear Energy Institute) has developed a guidance called 'NEI PRA Peer Review Guidance,' and NRC (Nuclear Regulatory Committee) and ASME have developed the 'PRA Standard.' In Korea, several projects are on going now, such as the extension of AOT/STI of RPS/ESFAS, Risk-Informed In-Service Inspection (RI-ISI). However, in Korea, there have been no attempts to evaluate the quality of PSA model itself. Therefore, we cannot be sure about the quality of PSA whether or not the present PSA model can be used for the risk-informed applications such as mentioned above. We can say that the evaluation of PSA model quality is the basis for the RIPBO. In this report, we have evaluated the quality of PSA model for Ulchin 3 and 4 units based on the NEI guidance. We, also, have derived what items are to be improved to upgrade the quality of PSA model and how it can be improved. This report can be used as the base of RIPBO work in Korea. The review result based on ASME Standard is published as the separated technical report of KAERI

Assessing the accuracy of remote sensing techniques in vegetation ...

African Journals Online (AJOL)

Assessing the accuracy of remote sensing techniques in vegetation fractions estimation. ... This study aimed at exploring different remote sensing (RS) techniques for quantitatively measuring vegetation and bare soil ... HOW TO USE AJOL.

International Space Station End-of-Life Probabilistic Risk Assessment

Science.gov (United States)

Duncan, Gary W.

2014-01-01

The International Space Station (ISS) end-of-life (EOL) cycle is currently scheduled for 2020, although there are ongoing efforts to extend ISS life cycle through 2028. The EOL for the ISS will require deorbiting the ISS. This will be the largest manmade object ever to be de-orbited therefore safely deorbiting the station will be a very complex problem. This process is being planned by NASA and its international partners. Numerous factors will need to be considered to accomplish this such as target corridors, orbits, altitude, drag, maneuvering capabilities etc. The ISS EOL Probabilistic Risk Assessment (PRA) will play a part in this process by estimating the reliability of the hardware supplying the maneuvering capabilities. The PRA will model the probability of failure of the systems supplying and controlling the thrust needed to aid in the de-orbit maneuvering.

Nasal PCR assay for the detection of Mycobacterium leprae pra gene to study subclinical infection in a community.

Science.gov (United States)

Arunagiri, Kamalanathan; Sangeetha, Gopalakrishnan; Sugashini, Padmavathy Krishnan; Balaraman, Sekar; Showkath Ali, M K

2017-03-01

Leprosy is a chronic infectious disease caused by Mycobacterium leprae. Identification of Mycobacterium leprae is difficult in part due to the inability of the leprosy bacillus to grow in vitro. A number of diagnostic methods for leprosy diagnosis have been proposed. Both serological tests and molecular probes have shown certain potential for detection and identification of Mycobacterium leprae in patients. In this study, we have investigated whether Mycobacterium leprae DNA from the nasal secretion of healthy household contacts and the non contacts could be detected through PCR amplification as a method to study the sub clinical infection in a community. A total of 200 samples, 100 each from contacts and non contacts representing all age groups and sex were included in this study. The M. leprae specific primer (proline-rich region) of pra gene was selected and PCR was performed using extracted DNA from the sample. A total of 13 samples were found to be positive for nasal PCR for pra gene among the male and female contacts out of which 7% were males and 6% were females. Even though several diagnostic tools are available to detect the cases of leprosy, they lack the specificity and sensitivity. PCR technology has demonstrated the improved diagnostic accuracy for epidemiological studies and requires minimal time. Although nasal PCR studies have been reported from many countries it is not usually recommended due to the high percentage of negative results in the contact. Copyright © 2017 Elsevier Ltd. All rights reserved.

Towards ABET accreditation for a SWE program: alternative student assessment techniques

International Nuclear Information System (INIS)

Alghamdi, A.; Nasir, M.; Alnafjan, K.

2011-01-01

This paper describes assessment techniques utilized for assessing undergraduate students studying in a software engineering program. The purpose behind this work is to get the program accredited by the Accreditation Board of Engineering and Technology (ABET). Therefore, a number of applied direct and indirect assessment techniques are described. These techniques are implemented towards the end of the semester to assess the extent to which the student and course outcomes are satisfied. Consequently, results are obtained and analyzed and various learning issues are eventually identified. Finally, the paper provides suggestions for improvement in course delivery as well as learning mechanism. (author)

Probabilistic risk assessment in the nuclear power industry

International Nuclear Information System (INIS)

Fullwood, R.R.; Hall, R.E.

1988-01-01

This book describes the more important improvements in risk assessment methodology developed over the last decade. The book covers the following areas - a general view of risk pertaining to nuclear power, mathematics necessary to understand the text, a concise overview of the light water reactors and their features for protecting the public, probabilities and consequences calculated to form risk assessment to the plant, and 34 applications of probabilistic risk assessment (PRA) in the power generation industry. There is a glossary of acronyms and unusual words and a list of references. (author)

Risk assessment techniques for civil aviation security

Energy Technology Data Exchange (ETDEWEB)

Tamasi, Galileo, E-mail: g.tamasi@enac.rupa.i [Ente Nazionale per l' Aviazione Civile-Direzione Progetti, Studi e Ricerche, Via di Villa Ricotti, 42, 00161 Roma (Italy); Demichela, Micaela, E-mail: micaela.demichela@polito.i [SAfeR-Centro Studi su Sicurezza, Affidabilita e Rischi, Dipartimento di Scienza dei Materiali e Ingegneria Chimica, Politecnico di Torino, Corso Duca degli Abruzzi, 24, 10129 Torino (Italy)

2011-08-15

Following the 9/11 terrorists attacks in New York a strong economical effort was made to improve and adapt aviation security, both in infrastructures as in airplanes. National and international guidelines were promptly developed with the objective of creating a security management system able to supervise the identification of risks and the definition and optimization of control measures. Risk assessment techniques are thus crucial in the above process, since an incorrect risk identification and quantification can strongly affect both the security level as the investments needed to reach it. The paper proposes a set of methodologies to qualitatively and quantitatively assess the risk in the security of civil aviation and the risk assessment process based on the threats, criticality and vulnerabilities concepts, highlighting their correlation in determining the level of risk. RAMS techniques are applied to the airport security system in order to analyze the protection equipment for critical facilities located in air-side, allowing also the estimation of the importance of the security improving measures vs. their effectiveness.

Risk assessment techniques for civil aviation security

International Nuclear Information System (INIS)

Tamasi, Galileo; Demichela, Micaela

2011-01-01

Following the 9/11 terrorists attacks in New York a strong economical effort was made to improve and adapt aviation security, both in infrastructures as in airplanes. National and international guidelines were promptly developed with the objective of creating a security management system able to supervise the identification of risks and the definition and optimization of control measures. Risk assessment techniques are thus crucial in the above process, since an incorrect risk identification and quantification can strongly affect both the security level as the investments needed to reach it. The paper proposes a set of methodologies to qualitatively and quantitatively assess the risk in the security of civil aviation and the risk assessment process based on the threats, criticality and vulnerabilities concepts, highlighting their correlation in determining the level of risk. RAMS techniques are applied to the airport security system in order to analyze the protection equipment for critical facilities located in air-side, allowing also the estimation of the importance of the security improving measures vs. their effectiveness.

Specificity and sensitivity assessment of selected nasal provocation testing techniques

Directory of Open Access Journals (Sweden)

Edyta Krzych-Fałta

2016-12-01

Full Text Available Introduction: Nasal provocation testing involves an allergen-specific local reaction of the nasal mucosa to the administered allergen. Aim: To determine the most objective nasal occlusion assessment technique that could be used in nasal provocation testing. Material and methods : A total of 60 subjects, including 30 patients diagnosed with allergy to common environmental allergens and 30 healthy subjects were enrolled into the study. The method used in the study was a nasal provocation test with an allergen, with a standard dose of a control solution and an allergen (5,000 SBU/ml administered using a calibrated atomizer into both nostrils at room temperature. Early-phase nasal mucosa response in the early phase of the allergic reaction was assessed via acoustic rhinometry, optical rhinometry, nitric oxide in nasal air, and tryptase levels in the nasal lavage fluid. Results : In estimating the homogeneity of the average values, the Levene’s test was used and receiver operating characteristic curves were plotted for all the methods used for assessing the nasal provocation test with an allergen. Statistically significant results were defined for p < 0.05. Of all the objective assessment techniques, the most sensitive and characteristic ones were the optical rhinometry techniques (specificity = 1, sensitivity = 1, AUC = 1, PPV = 1, NPV = 1. Conclusions : The techniques used showed significant differences between the group of patients with allergic rhinitis and the control group. Of all the objective assessment techniques, those most sensitive and characteristic were the optical rhinometry.

Coping with uncertainty in environmental impact assessments: Open techniques

Energy Technology Data Exchange (ETDEWEB)

Cardenas, Ibsen C., E-mail: c.cardenas@utwente.nl [IceBridge Research Institutea, Universiteit Twente, P.O. Box 217, 7500 AE Enschede (Netherlands); Halman, Johannes I.M., E-mail: J.I.M.Halman@utwente.nl [Universiteit Twente, P.O. Box 217, 7500 AE Enschede (Netherlands)

2016-09-15

Uncertainty is virtually unavoidable in environmental impact assessments (EIAs). From the literature related to treating and managing uncertainty, we have identified specific techniques for coping with uncertainty in EIAs. Here, we have focused on basic steps in the decision-making process that take place within an EIA setting. More specifically, we have identified uncertainties involved in each decision-making step and discussed the extent to which these can be treated and managed in the context of an activity or project that may have environmental impacts. To further demonstrate the relevance of the techniques identified, we have examined the extent to which the EIA guidelines currently used in Colombia consider and provide guidance on managing the uncertainty involved in these assessments. Some points that should be considered in order to provide greater robustness in impact assessments in Colombia have been identified. These include the management of stakeholder values, the systematic generation of project options, and their associated impacts as well as the associated management actions, and the evaluation of uncertainties and assumptions. We believe that the relevant and specific techniques reported here can be a reference for future evaluations of other EIA guidelines in different countries. - Highlights: • uncertainty is unavoidable in environmental impact assessments, EIAs; • we have identified some open techniques to EIAs for treating and managing uncertainty in these assessments; • points for improvement that should be considered in order to provide greater robustness in EIAs in Colombia have been identified; • the paper provides substantiated a reference for possible examinations of EIAs guidelines in other countries.

Coping with uncertainty in environmental impact assessments: Open techniques

International Nuclear Information System (INIS)

Cardenas, Ibsen C.; Halman, Johannes I.M.

2016-01-01

Uncertainty is virtually unavoidable in environmental impact assessments (EIAs). From the literature related to treating and managing uncertainty, we have identified specific techniques for coping with uncertainty in EIAs. Here, we have focused on basic steps in the decision-making process that take place within an EIA setting. More specifically, we have identified uncertainties involved in each decision-making step and discussed the extent to which these can be treated and managed in the context of an activity or project that may have environmental impacts. To further demonstrate the relevance of the techniques identified, we have examined the extent to which the EIA guidelines currently used in Colombia consider and provide guidance on managing the uncertainty involved in these assessments. Some points that should be considered in order to provide greater robustness in impact assessments in Colombia have been identified. These include the management of stakeholder values, the systematic generation of project options, and their associated impacts as well as the associated management actions, and the evaluation of uncertainties and assumptions. We believe that the relevant and specific techniques reported here can be a reference for future evaluations of other EIA guidelines in different countries. - Highlights: • uncertainty is unavoidable in environmental impact assessments, EIAs; • we have identified some open techniques to EIAs for treating and managing uncertainty in these assessments; • points for improvement that should be considered in order to provide greater robustness in EIAs in Colombia have been identified; • the paper provides substantiated a reference for possible examinations of EIAs guidelines in other countries.

Probabilistic Risk Assessment for Decision Making During Spacecraft Operations

Science.gov (United States)

Meshkat, Leila

2009-01-01

Decisions made during the operational phase of a space mission often have significant and immediate consequences. Without the explicit consideration of the risks involved and their representation in a solid model, it is very likely that these risks are not considered systematically in trade studies. Wrong decisions during the operational phase of a space mission can lead to immediate system failure whereas correct decisions can help recover the system even from faulty conditions. A problem of special interest is the determination of the system fault protection strategies upon the occurrence of faults within the system. Decisions regarding the fault protection strategy also heavily rely on a correct understanding of the state of the system and an integrated risk model that represents the various possible scenarios and their respective likelihoods. Probabilistic Risk Assessment (PRA) modeling is applicable to the full lifecycle of a space mission project, from concept development to preliminary design, detailed design, development and operations. The benefits and utilities of the model, however, depend on the phase of the mission for which it is used. This is because of the difference in the key strategic decisions that support each mission phase. The focus of this paper is on describing the particular methods used for PRA modeling during the operational phase of a spacecraft by gleaning insight from recently conducted case studies on two operational Mars orbiters. During operations, the key decisions relate to the commands sent to the spacecraft for any kind of diagnostics, anomaly resolution, trajectory changes, or planning. Often, faults and failures occur in the parts of the spacecraft but are contained or mitigated before they can cause serious damage. The failure behavior of the system during operations provides valuable data for updating and adjusting the related PRA models that are built primarily based on historical failure data. The PRA models, in turn

Results of a nuclear power plant Application of a new technique for human error analysis (ATHEANA)

International Nuclear Information System (INIS)

Forester, J.A.; Whitehead, D.W.; Kolaczkowski, A.M.; Thompson, C.M.

1997-01-01

A new method to analyze human errors has been demonstrated at a pressurized water reactor (PWR) nuclear power plant. This was the first application of the new method referred to as A Technique for Human Error Analysis (ATHEANA). The main goals of the demonstration were to test the ATHEANA process as described in the frame-of-reference manual and the implementation guideline, test a training package developed for the method, test the hypothesis that plant operators and trainers have significant insight into the error-forcing-contexts (EFCs) that can make unsafe actions (UAs) more likely, and to identify ways to improve the method and its documentation. A set of criteria to evaluate the open-quotes successclose quotes of the ATHEANA method as used in the demonstration was identified. A human reliability analysis (HRA) team was formed that consisted of an expert in probabilistic risk assessment (PRA) with some background in HRA (not ATHEANA) and four personnel from the nuclear power plant. Personnel from the plant included two individuals from their PRA staff and two individuals from their training staff. Both individuals from training are currently licensed operators and one of them was a senior reactor operator open-quotes on shiftclose quotes until a few months before the demonstration. The demonstration was conducted over a 5 month period and was observed by members of the Nuclear Regulatory Commission's ATHEANA development team, who also served as consultants to the HRA team when necessary. Example results of the demonstration to date, including identified human failure events (HFEs), UAs, and EFCs are discussed. Also addressed is how simulator exercises are used in the ATHEANA demonstration project

Development of A Methodology for Assessing Various Accident Management Strategies Using Decision Tree Models

Energy Technology Data Exchange (ETDEWEB)

Kim, Nam Yeong; Kim, Jin Tae; Jae, Moo Sung [Hanyang University, Seoul (Korea, Republic of); Jerng, Dong Wook [Chung-Ang University, Seoul (Korea, Republic of)

2016-05-15

The purpose of ASP (Accident Sequence Precursor) analysis is to evaluate operational accidents in full power and low power operation by using PRA (Probabilistic Risk Assessment) technologies. The awareness of the importance of ASP analysis has been on rise. The methodology for ASP analysis has been developed in Korea, KINS (Korea Institute of Nuclear Safety) has managed KINS-ASP program since it was developed. In this study, we applied ASP analysis into operational accidents in full power and low power operation to quantify CCDP (Conditional Core Damage Probability). To reflect these 2 cases into PRA model, we modified fault trees and event trees of the existing PRA model. Also, we suggest the ASP regulatory system in the conclusion. In this study, we reviewed previous studies for ASP analysis. Based on it, we applied it into operational accidents in full power and low power operation. CCDP of these 2 cases are 1.195E-06 and 2.261E-03. Unlike other countries, there is no regulatory basis of ASP analysis in Korea. ASP analysis could detect the risk by assessing the existing operational accidents. ASP analysis can improve the safety of nuclear power plant by detecting, reviewing the operational accidents, and finally removing potential risk. Operator have to notify regulatory institute of operational accident before operator takes recovery work for the accident. After follow-up accident, they have to check precursors in data base to find similar accident.

Classroom Assessment Techniques

Science.gov (United States)

Narayanan, M.

2003-12-01

and the learner should be carefully observed and monitored. Forrest says that Student Portfolios, which document learning in more detail, seldom reveal how teaching contributes to students' progress. Cerbin further indicates that a course portfolio is essentially, a like a manuscript of scholarly work in progress. In this example, it is a work that explains what, how, and why students learn or do not learn in a course. In this paper, the author reports on a dozen techniques that could perhaps be used to document assessment of student learning. References : Cerbin, W. (1993). Fostering a culture of teaching as scholarship. The Teaching Professor, 7(3), 1-2. Edgerton, R., Hutchings, P., & Quinlan, P. (1991). The teaching portfolio: Capturing the scholarship in teaching. Washington, DC: American Association for Higher Education. Forrest, A. (1990). Time will tell: Portfolio-assisted assessment of general education. Washington, DC: American Association for Higher Education. Linn, R., Baker, E., & Dunbar, S. (1991). Complex, Performance-based Assessment: Expectations and Validation Criteria. Educational Researcher, 20 (8), 15-21. Narayanan, M. (2003). Assessment in Higher Education: Partnerships in Learning. Paper presented at the 23rd Annual Lilly Conference on College Teaching, Miami University, Oxford, OH. Seldin, P. (1991). The teaching portfolio. Bolton, MA: Anker. Young, C. O., Sr., & Young, L. H. (1999). Assessing Learning in Interactive Courses. Journal on Excellence in College Teaching, 10 (1), 63-76.

Impact of Gold mining activities on the water quality of the lower pra river

International Nuclear Information System (INIS)

Dwamena, Offei Samuel K.

2013-07-01

This study was conducted to assess the extent of Mercury (THg) contamination at four locations within the Shama-Mporhor Wassa catchment area of the Lower Pra River. Water, fish and sediment samples were taken twice with the longitudinal transect method at Daboase, Beposo, Bokorkope and Shama during the minor rainy season in October and at the apex of the dry season in March. Careful investigation of the Shama-Mporhor Wassa catchment area revealed that two of the locations Daboase and Beposo had been continuously impacted by the activities of Artisanal Gold miners (AGM). From the study, Total Mercury (THg) levels were found to have persisted in River water several kilometers downstream the second Artisanal Gold mining (AGM) location at Shama estuary for both seasons. Ten trace elements Mercury (Hg), Selenium (Se), Copper (Cu), Chromium (Cr), Lead (Pb), Iron (Fe), Manganese (Mn), Nickel (Ni), Zinc (Zn) and Cadmium (Cd) were determined in water, fish and sediment samples using the Atomic Absorption Spectroscopy (AAS) equipped with both Hydride Generation (HGAAS) for Selenium (Se) and Cold Vapour (CVAAS) for Total Mercury (THg). The levels of Total Mercury (THg) were largely above the WHO and USEPA guidelines for drinking water (1μg/L) and sediments (200 μg/Kg) respectively for the four locations investigated. Total Mercury (THg) exceeded the WHO, 2011 guideline value of 0.5 mg/Kg for fish species Clarias submarginatus but was below the guideline value for Xenomystus nigri. Mean concentration of Cd and Fe exceeded the WHO, 2011 guideline values for drinking water for the wet season. The other trace elements Zn, Ni, Cu, Cr, Se, Mn, and Pb had their mean concentration below the WHO, 2011 guideline values for drinking water. Apart from the mean concentration of Cd that exceeded the Canadian Interim Sediment Quality (ISQG) guideline value of 0.6 mg/Kg for the wet season, Cr, Cu, Zn, Ni and Pb were below their respective guideline values for both seasons. Statistical

Development of probabilistic risk assessment methodology against extreme snow for sodium-cooled fast reactor

Energy Technology Data Exchange (ETDEWEB)

Yamano, Hidemasa, E-mail: yamano.hidemasa@jaea.go.jp; Nishino, Hiroyuki; Kurisaka, Kenichi

2016-11-15

Highlights: • Snow PRA methodology was developed. • Snow hazard category was defined as the combination of daily snowfall depth (speed) and snowfall duration. • Failure probability models of snow removal action, manual operation of the air cooler dampers and the access route were developed. • Snow PRA showed less than 10{sup −6}/reactor-year of core damage frequency. - Abstract: This paper describes snow probabilistic risk assessment (PRA) methodology development through external hazard and event sequence evaluations mainly in terms of decay heat removal (DHR) function of a sodium-cooled fast reactor (SFR). Using recent 50-year weather data at a typical Japanese SFR site, snow hazard categories were set for the combination of daily snowfall depth (snowfall speed) and snowfall duration which can be calculated by dividing the snow depth by the snowfall speed. For each snow hazard category, the event sequence was evaluated by event trees which consist of several headings representing the loss of DHR. Snow removal action and manual operation of the air cooler dampers were introduced into the event trees as accident managements. Access route failure probability model was also developed for the quantification of the event tree. In this paper, the snow PRA showed less than 10{sup −6}/reactor-year of core damage frequency. The dominant snow hazard category was the combination of 1–2 m/day of snowfall speed and 0.5–0.75 day of snowfall duration. Importance and sensitivity analyses indicated a high risk contribution of the securing of the access routes.

Development of probabilistic risk assessment methodology against extreme snow for sodium-cooled fast reactor

International Nuclear Information System (INIS)

Yamano, Hidemasa; Nishino, Hiroyuki; Kurisaka, Kenichi

2016-01-01

Highlights: • Snow PRA methodology was developed. • Snow hazard category was defined as the combination of daily snowfall depth (speed) and snowfall duration. • Failure probability models of snow removal action, manual operation of the air cooler dampers and the access route were developed. • Snow PRA showed less than 10"−"6/reactor-year of core damage frequency. - Abstract: This paper describes snow probabilistic risk assessment (PRA) methodology development through external hazard and event sequence evaluations mainly in terms of decay heat removal (DHR) function of a sodium-cooled fast reactor (SFR). Using recent 50-year weather data at a typical Japanese SFR site, snow hazard categories were set for the combination of daily snowfall depth (snowfall speed) and snowfall duration which can be calculated by dividing the snow depth by the snowfall speed. For each snow hazard category, the event sequence was evaluated by event trees which consist of several headings representing the loss of DHR. Snow removal action and manual operation of the air cooler dampers were introduced into the event trees as accident managements. Access route failure probability model was also developed for the quantification of the event tree. In this paper, the snow PRA showed less than 10"−"6/reactor-year of core damage frequency. The dominant snow hazard category was the combination of 1–2 m/day of snowfall speed and 0.5–0.75 day of snowfall duration. Importance and sensitivity analyses indicated a high risk contribution of the securing of the access routes.

The tsunami probabilistic risk assessment of nuclear power plant (3). Outline of tsunami fragility analysis

International Nuclear Information System (INIS)

Mihara, Yoshinori

2012-01-01

Tsunami Probabilistic Risk Assessment (PRA) standard was issued in February 2012 by Standard Committee of Atomic Energy Society of Japan (AESJ). This article detailed tsunami fragility analysis, which calculated building and structure damage probability contributing core damage and consisted of five evaluation steps: (1) selection of evaluated element and damage mode, (2) selection of evaluation procedure, (3) evaluation of actual stiffness, (4) evaluation of actual response and (5) evaluation of fragility (damage probability and others). As an application example of the standard, calculation results of tsunami fragility analysis investigation by tsunami PRA subcommittee of AESJ were shown reflecting latest knowledge of damage state caused by wave force and others acted by tsunami from the 'off the Pacific Coast of Tohoku Earthquake'. (T. Tanaka)

SLIM-MAUD - a computer based technique for human reliability assessment

International Nuclear Information System (INIS)

Embrey, D.E.

1985-01-01

The Success Likelihood Index Methodology (SLIM) is a widely applicable technique which can be used to assess human error probabilities in both proceduralized and cognitive tasks (i.e. those involving decision making, problem solving, etc.). It assumes that expert assessors are able to evaluate the relative importance (or weights) of different factors called Performance Shaping Factors (PSFs), in determining the likelihood of error for the situations being assessed. Typical PSFs are the extent to which good procedures are available, operators are adequately trained, the man-machine interface is well designed, etc. If numerical ratings are made of the PSFs for the specific tasks being evaluated, these can be combined with the weights to give a numerical index, called the Success Likelihood Index (SLI). The SLI represents, in numerical form, the overall assessment of the experts of the likelihood of task success. The SLI can be subsequently transformed to a corresponding human error probability (HEP) estimate. The latest form of the SLIM technique is implemented using a microcomputer based system called MAUD (Multi-Attribute Utility Decomposition), the resulting technique being called SLIM-MAUD. A detailed description of the SLIM-MAUD technique and case studies of applications are available. An illustrative example of the application of SLIM-MAUD in probabilistic risk assessment is given

Probabilistic risk criteria and their application to nuclear chemical plant design

International Nuclear Information System (INIS)

Arthur, T.; Barnes, D.S.; Brown, M.L.; Taig, A.R.; Johnston, B.D.; Hayns, M.

1989-01-01

A nuclear chemical plant safety strategy is presented. The use of risk criteria in design is demonstrated by reference to a particular area of the plant. This involves the application of Probabilistic Risk Assessment (PRA) techniques. Computer programs developed by the UK Atomic Energy Authority (UKAEA) at its Safety and Reliability Directorate (SRD) are used toe valuate and analyze the resultant fault trees. the magnitude of releases are estimated and individual and societal risks determined. The paper concludes that the application of PRA to a nuclear chemical plant can be structured in such a way as to allow a designer to work to quantitative risk targets

Cardiovascular risk from water arsenic exposure in Vietnam: Application of systematic review and meta-regression analysis in chemical health risk assessment.

Science.gov (United States)

Phung, Dung; Connell, Des; Rutherford, Shannon; Chu, Cordia

2017-06-01

A systematic review (SR) and meta-analysis cannot provide the endpoint answer for a chemical risk assessment (CRA). The objective of this study was to apply SR and meta-regression (MR) analysis to address this limitation using a case study in cardiovascular risk from arsenic exposure in Vietnam. Published studies were searched from PubMed using the keywords of arsenic exposure and cardiovascular diseases (CVD). Random-effects meta-regression was applied to model the linear relationship between arsenic concentration in water and risk of CVD, and then the no-observable-adverse-effect level (NOAEL) were identified from the regression function. The probabilistic risk assessment (PRA) technique was applied to characterize risk of CVD due to arsenic exposure by estimating the overlapping coefficient between dose-response and exposure distribution curves. The risks were evaluated for groundwater, treated and drinking water. A total of 8 high quality studies for dose-response and 12 studies for exposure data were included for final analyses. The results of MR suggested a NOAEL of 50 μg/L and a guideline of 5 μg/L for arsenic in water which valued as a half of NOAEL and guidelines recommended from previous studies and authorities. The results of PRA indicated that the observed exposure level with exceeding CVD risk was 52% for groundwater, 24% for treated water, and 10% for drinking water in Vietnam, respectively. The study found that systematic review and meta-regression can be considered as an ideal method to chemical risk assessment due to its advantages to bring the answer for the endpoint question of a CRA. Copyright © 2017 Elsevier Ltd. All rights reserved.

Risk-assessment techniques and the reactor licensing process

International Nuclear Information System (INIS)

Levine, S.

1979-01-01

A brief description of the Reactor Safety Study (WASH-1400), concentrating on the engineering aspects of the contribution to reactor accident risks is followed by some comments on how we have applied the insights and techniques developed in this study to prepare a program to improve the safety of nuclear power plants. Some new work we are just beginning on the application of risk-assessment techniques to stablize the reactor licensing process is also discussed

Issues related to structural aging in probabilistic risk assessment of nuclear power plants

International Nuclear Information System (INIS)

Ellingwood, Bruce R.

1998-01-01

Structural components and systems have an important safety function in nuclear power plants. Although they are essentially passive under normal operating conditions, they play a key role in mitigating the impact of extreme environmental events such as earthquakes, winds, fire and floods on plant safety. Moreover, the importance of structural components and systems in accident mitigation is amplified by common-cause effects. Reinforced concrete structural components and systems in NPPs are subject to a phenomenon known as aging, leading to time-dependent changes in strength and stiffness that may impact their ability to withstand various challenges during their service lives from operation, the environment and accidents. Time-dependent changes in structural properties as well as challenges to the system are random in nature. Accordingly, condition assessment of existing structures should be performed within a probabilistic framework. The mathematical formalism of a probabilistic risk assessment (PRA) provides a means for identifying aging structural components that may play a significant role in mitigating plant risk. Structural condition assessments supporting a decision regarding continued service can be rendered more efficient if guided by the logic of a PRA

Doppler ultrasound imaging techniques for assessment of synovial inflammation

Directory of Open Access Journals (Sweden)

Filippucci E

2013-09-01

Full Text Available Emilio Filippucci,1 Fausto Salaffi,1 Marina Carotti,2 Walter Grassi1 1Rheumatology Department, Polytechnic University of the Marche, Ancona, Italy; 2Department of Radiology, Polytechnic University of the Marche, Ancona, Italy Abstract: Ultrasound is an evolving technique, and the rapid progress made in ultrasound technology over the past ten years has dramatically increased its range of applications in rheumatology. One of the most exciting advances is the use of Doppler ultrasound imaging in the assessment of blood flow abnormalities at the synovial tissue level in patients with chronic inflammatory arthritis. This review describes the Doppler techniques available and their main applications in patients with inflammatory arthritis, discusses the evidence supporting their use, and outlines the latest advances in hardware and software. Spectral, color, and power Doppler allow sensitive assessment of vascular abnormalities at the synovial tissue level. Use of contrast agents enhances visualization of the small synovial vessels using color or power Doppler ultrasound and allows for accurate characterization of the rheumatoid pannus. Doppler techniques represent a unique method for assessment of synovial inflammation, showing blood flow characteristics in real time. They are safe, noninvasive, cost-effective, and have high sensitivity in revealing and monitoring synovitis. However, several questions still need to be answered. In the near future, the Doppler techniques described here, together with upcoming hardware and software facilities, will be investigated further and a consensus will be reached on their feasibility and appropriate use in daily rheumatologic practice. Keywords: power and color Doppler techniques, ultrasound, contrast media, synovitis, rheumatoid arthritis

A socio-technical, probabilistic risk assessment model for surgical site infections in ambulatory surgery centers.

Science.gov (United States)

Bish, Ebru K; El-Amine, Hadi; Steighner, Laura A; Slonim, Anthony D

2014-10-01

To understand how structural and process elements may affect the risk for surgical site infections (SSIs) in the ambulatory surgery center (ASC) environment, the researchers employed a tool known as socio-technical probabilistic risk assessment (ST-PRA). ST-PRA is particularly helpful for estimating risks in outcomes that are very rare, such as the risk of SSI in ASCs. Study objectives were to (1) identify the risk factors associated with SSIs resulting from procedures performed at ASCs and (2) design an intervention to mitigate the likelihood of SSIs for the most common risk factors that were identified by the ST-PRA for a particular surgical procedure. ST-PRA was used to study the SSI risk in the ASC setting. Both quantitative and qualitative data sources were utilized, and sensitivity analysis was performed to ensure the robustness of the results. The event entitled "fail to protect the patient effectively" accounted for 51.9% of SSIs in the ambulatory care setting. Critical components of this event included several failure risk points related to skin preparation, antibiotic administration, staff training, proper response to glove punctures during surgery, and adherence to surgical preparation rules related to the wearing of jewelry, watches, and artificial nails. Assuming a 75% reduction in noncompliance on any combination of 2 of these 5 components, the risk for an SSI decreased from 0.0044 to between 0.0027 and 0.0035. An intervention that targeted the 5 major components of the major risk point was proposed, and its implications were discussed.

The Environmental Assessment Technique: An Empirical Study.

Science.gov (United States)

Overall, Jesse U., IV

The purpose of the study was to investigate the effectiveness of Alexander Astin's Environmental Assessment Technique (EAT) in describing the environmental press at a large public university, California State University at Los Angeles. Results indicate that EAT is a very economical method for broadly describing aspects of a university's…

Probabilistic risk analysis for the NASA space shuttle: a brief history and current work

International Nuclear Information System (INIS)

Pate-Cornell, Elisabeth; Dillon, Robin

2001-01-01

While NASA managers have always relied on risk analysis tools for the development and maintenance of space projects, quantitative and especially probabilistic techniques have been gaining acceptance in recent years. In some cases, the studies have been required, for example, to launch the Galileo spacecraft with plutonium fuel, but these successful applications have helped to demonstrate the benefits of these tools. This paper reviews the history of probabilistic risk analysis (PRA) by NASA for the space shuttle program and discusses the status of the on-going development of the Quantitative Risk Assessment System (QRAS) software that performs PRA. The goal is to have within NASA a tool that can be used when needed to update previous risk estimates and to assess the benefits of possible upgrades to the system

Methods for the elicitation and use of expert opinion in risk assessment: Phase 1, A critical evaluation and directions for future research

International Nuclear Information System (INIS)

Mosleh, A.; Bier, V.M.; Apostolakis, G.

1987-08-01

The purpose of this work is to critically review and evaluate the elicitation and use of expert opinion in probabilistic risk assessment (PRA) in light of the available empirical and theoretical results on expert opinion use. PRA practice is represented by five case studies selected to represent a variety of aspects of the problem: assessments of component failure rates and maintenance data; recent assessments of seismic hazard rates; assessments of containment phenomenology; assessments of human error rates; and accident precursor studies. The review has yielded mixed results. On the negative side, there appears to be little reliance on normative expertise in structuring the process of expert opinion elicitation and use; most applications instead rely primarily on the common sense of the experts involved in the analysis, which is not always an adequate guide. On the positive side, however, there is evidence that expert opinions can in fact be used well in practical settings. Suggestions are given for Phase II work to enhance the applicability and use of appropriate expert opinion methods

Recommendations for a proposed standard for performing systems analysis

International Nuclear Information System (INIS)

LaChance, J.; Whitehead, D.; Drouin, M.

1998-01-01

In August 1995, the Nuclear Regulatory Commission (NRC) issued a policy statement proposing improved regulatory decisionmaking by increasing the use of PRA [probabilistic risk assessment] in all regulatory matters to the extent supported by the state-of-the-art in PRA methods and data. A key aspect in using PRA in risk-informed regulatory activities is establishing the appropriate scope and attributes of the PRA. In this regard, ASME decided to develop a consensus PRA Standard. The objective is to develop a PRA Standard such that the technical quality of nuclear plant PRAs will be sufficient to support risk-informed regulatory applications. This paper presents examples recommendations for the systems analysis element of a PRA for incorporation into the ASME PRA Standard

International status of application of probabilistic risk analysis

International Nuclear Information System (INIS)

Cullingford, M.C.

1984-01-01

Probabilistic Risk Assessment (PRA) having been practised for about ten years and with more than twenty studies completed has reached a level of maturity such that the insights and other products derived from specific studies may be assessed. The first full-scale PRA studies were designed to develop the methodology and assess the overall risk from nuclear power. At present PRA is performed mostly for individual plants to identify core damage accident sequences and significant contributors to such sequences. More than 25 countries are utilizing insights from PRA, some from full-scale PRA studies and other countries by performing reliability analyses on safety systems identified as important contributors to one or more core melt sequences. Many Member States of the IAEA fall into one of three groups: those having (a) a large, (b) a medium number of reactor-years of operating experience and (c) those countries in the planning or feasibility study stages of a nuclear power programme. Of the many potential uses of PRA the decision areas of safety improvement by backfitting, development of operating procedures and as the basis of standards are felt to be important by countries of all three groups. The use of PRA in showing compliance with safety goals and for plant availability studies is held to be important only by those countries which have operating experience. The evolution of the PRA methodology has led to increased attention to quantification of uncertainties both in the probabilities and consequences. Although many products from performing a PRA do not rely upon overall risk numbers, increasing emphasis is being placed on the interpretation of uncertainties in risk numbers for use in decisions. International co-operation through exchange of information regarding experience with PRA methodology and its application to nuclear safety decisions will greatly enhance the widespread use of PRA. (author)

Exploratory Nuclear Reactor Safety Analysis and Visualization via Integrated Topological and Geometric Techniques

Energy Technology Data Exchange (ETDEWEB)

Maljovec, Dan [Univ. of Utah, Salt Lake City, UT (United States); Wang, Bei [Univ. of Utah, Salt Lake City, UT (United States); Pascucci, Valerio [Univ. of Utah, Salt Lake City, UT (United States); Bremer, Peer-Timo [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Mandelli, Diego [Idaho National Lab. (INL), Idaho Falls, ID (United States); Pernice, Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nourgaliev, Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2013-10-01

A recent trend in the nuclear power engineering field is the implementation of heavily computational and time consuming algorithms and codes for both design and safety analysis. In particular, the new generation of system analysis codes aim to embrace several phenomena such as thermo-hydraulic, structural behavior, and system dynamics, as well as uncertainty quantification and sensitivity analyses. The use of dynamic probabilistic risk assessment (PRA) methodologies allows a systematic approach to uncertainty quantification. Dynamic methodologies in PRA account for possible coupling between triggered or stochastic events through explicit consideration of the time element in system evolution, often through the use of dynamic system models (simulators). They are usually needed when the system has more than one failure mode, control loops, and/or hardware/process/software/human interaction. Dynamic methodologies are also capable of modeling the consequences of epistemic and aleatory uncertainties. The Monte-Carlo (MC) and the Dynamic Event Tree (DET) approaches belong to this new class of dynamic PRA methodologies. The major challenges in using MC and DET methodologies (as well as other dynamic methodologies) are the heavier computational and memory requirements compared to the classical ET analysis. This is due to the fact that each branch generated can contain time evolutions of a large number of variables (about 50,000 data channels are typically present in RELAP) and a large number of scenarios can be generated from a single initiating event (possibly on the order of hundreds or even thousands). Such large amounts of information are usually very difficult to organize in order to identify the main trends in scenario evolutions and the main risk contributors for each initiating event. This report aims to improve Dynamic PRA methodologies by tackling the two challenges mentioned above using: 1) adaptive sampling techniques to reduce computational cost of the analysis

Application of fisheries management techniques to assessing impacts: task I report

International Nuclear Information System (INIS)

McKenzie, D.H.; Baker, K.S.; Fickeisen, D.H.; Metzger, R.M.; Skalski, J.R.

1979-03-01

Task I efforts examined the available fisheries management techniques and assessed their potential application in a confirmatory monitoring program. The objective of such monitoring programs is to confirm that the prediction of an insignificant impact (usually made in the FES) was correct. Fisheries resource managers have developed several tools for assessing the fish population response to stress (exploitation) and they were thought potentially useful for detecting nuclear power plant impacts. Techniques in three categories were examined; catch removal, population dynamics, and nondestructive censuses, and the report contains their description, examples of application, advantages, and disadvantages. The techniques applied at nuclear power plant sites were examined in detail to provide information on implementation and variability of specific approaches. The most suitable techniques to incorporate into a monitoring program confirming no impact appear to be those based on Catch Per Unity Effort (CPUE) and hydroacoustic data. In some specific cases, age and growth studies and indirect census techniques may be beneficial. Recommendations for task II efforts to incorporate these techniques into monitoring program designs are presented. These include development of guidelines for; (1) designing and implementing a data collection program; (2) interpreting these data and assessing the occurrence of impact, and (3) establishment of the monitoring program's ability to detect changes in the affected populations

An uncertainty inventory demonstration - a primary step in uncertainty quantification

Energy Technology Data Exchange (ETDEWEB)

Langenbrunner, James R. [Los Alamos National Laboratory; Booker, Jane M [Los Alamos National Laboratory; Hemez, Francois M [Los Alamos National Laboratory; Salazar, Issac F [Los Alamos National Laboratory; Ross, Timothy J [UNM

2009-01-01

Tools, methods, and theories for assessing and quantifying uncertainties vary by application. Uncertainty quantification tasks have unique desiderata and circumstances. To realistically assess uncertainty requires the engineer/scientist to specify mathematical models, the physical phenomena of interest, and the theory or framework for assessments. For example, Probabilistic Risk Assessment (PRA) specifically identifies uncertainties using probability theory, and therefore, PRA's lack formal procedures for quantifying uncertainties that are not probabilistic. The Phenomena Identification and Ranking Technique (PIRT) proceeds by ranking phenomena using scoring criteria that results in linguistic descriptors, such as importance ranked with words, 'High/Medium/Low.' The use of words allows PIRT to be flexible, but the analysis may then be difficult to combine with other uncertainty theories. We propose that a necessary step for the development of a procedure or protocol for uncertainty quantification (UQ) is the application of an Uncertainty Inventory. An Uncertainty Inventory should be considered and performed in the earliest stages of UQ.

Expertise, Domains, and the Consensual Assessment Technique

Science.gov (United States)

Kaufman, James C.; Baer, John; Cole, Jason C.

2009-01-01

The Consensual Assessment Technique (CAT) argues that the most valid judgments of the creativity are those of the combined opinions of experts in the field. Yet who exactly qualifies as an expert to evaluate a creative product such as a short story? This study examines both novice and expert judgments of student short fiction. Results indicate a…

Assessment of hand blood flow: a modified technique

International Nuclear Information System (INIS)

Kirsh, J.C.; Tepperman, P.S.

1985-01-01

A blood flow artifact has been identified with the conventional bolus-injection technique in radionuclide studies of hand disorders. The artifact, consisting of increased blood flow on the injected side, was demonstrated in 22 of 25 subjects. Using a modified injection technique to allow time for local blood flow to return to the basal state, the artifact could be eliminated in 19 of 23 additional subjects. Use of this simple protocol should help avoid misinterpretation of blood flow asymmetry in the assessment of hand disorders

MENELAAH HUKUM WARIS PRA-ISLAM DAN AWAL ISLAM SERTA PELETAKAN DASAR-DASAR HUKUM KEWARISAN ISLAM

Directory of Open Access Journals (Sweden)

Asrizal Saiin

2017-03-01

[Dalam sejarahnya, pembagian harta warisan sudah ada sebelum Islam (pra-Islam. Adapun sistem pewarisannya adalah sistem keturunan dan sistem sebab. Pembagian harta warisan bersifat patrilinear, artinya anak-anak yang belum dewasa dan kaum perempuan tidak berhak mendapatkan harta warisan, sekalipun mereka merupakan ahli waris dari yang telah meninggal. Seseorang baru bisa mendapatkan harta apabila; adanya pertalian kerabat, janji ikatan prasetia, dan pengangkatan anak. Sementara pada masa awal Islam seseorang bisa mendapatkan harta warisan apabila; adanya pertalian kerabat, pengangkatan anak, adanya hijrah dan adanya persaudaraan. Pewarisan, baru terjadi jika ada sebab-sebab yang mengikat pewaris dengan ahli warisnya, seperti adanya perkawinan, kekerabatan, dan wala’. Adapun hal-hal yang dapat menggugurkan hak seseorang menerima warisan adalah; perbudakan, pembunuhan, berlainan agama, murtad, karena hilang tanpa berita dan berlainan negara. Sebelum pembagian warisan ada beberapa hak yang harus dipenuhi terlebih dahulu, seperti; hak yang berkaitan dengan zat harta peninggalan, biaya perawatan jenazah, pelunasan hutang dan pemberian wasiat

The verification tests of residual radioactivity measurement and assessment techniques for buildings and soils

International Nuclear Information System (INIS)

Onozawa, T.; Ishikura, T.; Yoshimura, Yukio; Nakazawa, M.; Makino, S.; Urayama, K.; Kawasaki, S.

1996-01-01

According to the standard procedure for decommissioning a commercial nuclear power plant (CNPP) in Japan, controlled areas will be released for unrestricted use before the dismantling of a reactor building. If manual survey and sampling techniques were applied to measurement for unrestricted release on and in the extensive surface of the building, much time and much specialized labor would be required to assess the appropriateness of the releasing. Therefore the authors selected the following three techniques for demonstrating reliability and applicability of the techniques for CNPPs: (1) technique of assessing radioactive concentration distribution on the surface of buildings (ADB); (2) technique of assessing radioactive permeation distribution in the concrete structure of buildings (APB); (3) technique of assessing radioactive concentration distribution in soil (ADS). These tests include the techniques of measuring and assessing very low radioactive concentration distribution on the extensive surfaces of buildings and the soil surrounding of a plant with automatic devices. Technical investigation and preliminary study of the verification tests were started in 1990. In the study, preconditions were clarified for each technique and the performance requirements were set up. Moreover, simulation models have been constructed for several feasible measurement method to assess their performance in terms of both measurement test and simulation analysis. Fundamental tests have been under way using small-scale apparatuses since 1994

ANALISIS INFLASI PRA DAN PASCA KRISIS MONETER DALAM PEREKONOMIAN INDONESIA

Directory of Open Access Journals (Sweden)

Deni Sri Haryati

2014-10-01

Full Text Available Abstrak ___________________________________________________________________ Penelitian ini menganalisis fluktuasi inflasi yang terjadi pada masa pra, saat, dan pasca krisis moneter. Terdapat 5 variabel yang berkorelasi dengan inflasi di Indonesia. Variabel-variabel tersebut adalah jumlah uang beredar (JUB, nilai tukar nominal (NTN, pertumbuhan ekonomi, pengeluaran pemerintah (PeP,dan  bahan bakar minyak (BBM. Sebenarnya variabel tersebut memiliki hubungan dengan inflasi pada seluruh era. Namun, variabel tesebut memiliki dominasi yang berbeda pada masing-masing eranya. Era prakrisis didominasi oleh 3 variabel, yakni; pertumbuhan ekonomi, NTN, dan JUB, era krisis didominasi oleh seluruh variabel yang ada, dan era pasca krisis didominasi oleh 3 variabel, yakni; PeP, NTN, dan JUB. Variabel-variabel ini akan dijelaskan pengaruhnya terhadap inflasi pada era yang didominasi dimana apakah memiliki hubungan sebanding atau berbanding terbalik. Abstract ___________________________________________________________________ This study analyzes the fluctuations in inflation that occurred in the pre, during and post the financial crisis. There are 5 variables that correlated with inflation in Indonesia. These variables are the money supply (MS, the nominal exchange rate (NER, economic growth, government expenditure (GE, and fuel oil (BBM. Actually, these variables have a relationship with inflation on the whole era. However, the variable has a predominance of different proficiency level in each era. Pre-crisis era dominated by three variables, namely; economic growth, NER, and MS, crisis era dominated by all the variables that exist, and the post-crisis era dominated by three variables, namely; GE, NER, and MS. These variables will be explained influence on inflation in an era dominated where or whether proportional or inversely proportional relationship. © 2014 Universitas Negeri Semarang

Applications of probabilistic risk analysis in nuclear criticality safety design

International Nuclear Information System (INIS)

Chang, J.K.

1992-01-01

Many documents have been prepared that try to define the scope of the criticality analysis and that suggest adding probabilistic risk analysis (PRA) to the deterministic safety analysis. The report of the US Department of Energy (DOE) AL 5481.1B suggested that an accident is credible if the occurrence probability is >1 x 10 -6 /yr. The draft DOE 5480 safety analysis report suggested that safety analyses should include the application of methods such as deterministic safety analysis, risk assessment, reliability engineering, common-cause failure analysis, human reliability analysis, and human factor safety analysis techniques. The US Nuclear Regulatory Commission (NRC) report NRC SG830.110 suggested that major safety analysis methods should include but not be limited to risk assessment, reliability engineering, and human factor safety analysis. All of these suggestions have recommended including PRA in the traditional criticality analysis

System Analysis and Risk Assessment system (SARA) Version 4.0

International Nuclear Information System (INIS)

Sattison, M.B.; Russell, K.D.; Skinner, N.L.

1992-01-01

This NUREG is the tutorial for the System Analysis and Risk Assessment System (SARA) Version 4.0, a microcomputer-based system used to analyze the safety issues of a family [i.e., a power plant, a manufacturing facility, any facility on which a probabilistic risk assessment (PRA) might be performed]. A series of lessons are provided that walk the user through some basic steps common to most analyses performed with SARA. The example problems presented in the lessons build on one another, and in combination, lead the user through all aspects of SARA sensitivity analysis

Probabilistic risk assessment for back-end facilities: Improving the treatment of fire and explosion scenarios

International Nuclear Information System (INIS)

Sunman, C.R.J.; Campbell, R.J.; Wakem, M.J.

1996-01-01

The nuclear reprocessing facilities at Sellafield are a key component of the International business of BNFL. The operations carried out at the site extend from the receipt and storage of irradiated fuel, chemical reprocessing, plutonium and uranium finishing, through mixed oxide fuel production. Additionally there are a wide range of supporting processes including solid waste encapsulation, vitrification, liquid waste evaporation and treatment. Decommissioning of the site's older facilities is also proceeding. The comprehensive range of these activities requires that the safety assessment team keeps up to date with developments in the field, as well as conducting and sponsoring appropriate research into methodologies and modelling in order to deliver a cost effective, timely service. This paper will review the role of Probabilistic Risk Assessment (PRA) in safety cases for operations at Sellafield and go on to describe some areas of PRA methodology development in the UK and in which BNFL is a contributor. Finally the paper will summarise some specific areas of methodology development associated with improving the modelling of fire and explosion hazards which are specific to BNFL. (author)

Probabilistic risk assessment modeling of digital instrumentation and control systems using two dynamic methodologies

Energy Technology Data Exchange (ETDEWEB)

Aldemir, T., E-mail: aldemir.1@osu.ed [Ohio State University, Nuclear Engineering Program, Columbus, OH 43210 (United States); Guarro, S. [ASCA, Inc., 1720 S. Catalina Avenue, Suite 220, Redondo Beach, CA 90277-5501 (United States); Mandelli, D. [Ohio State University, Nuclear Engineering Program, Columbus, OH 43210 (United States); Kirschenbaum, J. [Ohio State University, Department of Computer Science and Engineering, Columbus, OH 43210 (United States); Mangan, L.A. [Ohio State University, Nuclear Engineering Program, Columbus, OH 43210 (United States); Bucci, P. [Ohio State University, Department of Computer Science and Engineering, Columbus, OH 43210 (United States); Yau, M. [ASCA, Inc., 1720 S. Catalina Avenue, Suite 220, Redondo Beach, CA 90277-5501 (United States); Ekici, E. [Ohio State University, Department of Electrical and Computer Engineering, Columbus, OH 43210 (United States); Miller, D.W.; Sun, X. [Ohio State University, Nuclear Engineering Program, Columbus, OH 43210 (United States); Arndt, S.A. [U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001 (United States)

2010-10-15

The Markov/cell-to-cell mapping technique (CCMT) and the dynamic flowgraph methodology (DFM) are two system logic modeling methodologies that have been proposed to address the dynamic characteristics of digital instrumentation and control (I and C) systems and provide risk-analytical capabilities that supplement those provided by traditional probabilistic risk assessment (PRA) techniques for nuclear power plants. Both methodologies utilize a discrete state, multi-valued logic representation of the digital I and C system. For probabilistic quantification purposes, both techniques require the estimation of the probabilities of basic system failure modes, including digital I and C software failure modes, that appear in the prime implicants identified as contributors to a given system event of interest. As in any other system modeling process, the accuracy and predictive value of the models produced by the two techniques, depend not only on the intrinsic features of the modeling paradigm, but also and to a considerable extent on information and knowledge available to the analyst, concerning the system behavior and operation rules under normal and off-nominal conditions, and the associated controlled/monitored process dynamics. The application of the two methodologies is illustrated using a digital feedwater control system (DFWCS) similar to that of an operating pressurized water reactor. This application was carried out to demonstrate how the use of either technique, or both, can facilitate the updating of an existing nuclear power plant PRA model following an upgrade of the instrumentation and control system from analog to digital. Because of scope limitations, the focus of the demonstration of the methodologies was intentionally limited to aspects of digital I and C system behavior for which probabilistic data was on hand or could be generated within the existing project bounds of time and resources. The data used in the probabilistic quantification portion of the

NRC Support for the Kalinin (VVER) probabilistic risk assessment

International Nuclear Information System (INIS)

Bley, D.; Diamond, D.J.; Chu, T.L.; Azarm, A.; Pratt, W.T.; Johnson, D.; Szukiewicz, A.; Drouin, M.; El-Bassioni, A.; Su, T.M.

1998-01-01

The US Nuclear Regulatory Commission (NRC) and the Federal Nuclear and Radiation Safety Authority of the Russian Federation have been working together since 1994 to carry out a probabilistic risk assessment (PRA) of a VVER-1000 in the Russian Federation. This was a recognition by both parties that this technology has had a profound effect on the discipline of nuclear reactor safety in the West and that the technology should be transferred to others so that it can be applied to Soviet-designed plants. The NRC provided funds from the Agency for International Development and technical support primarily through Brookhaven National Laboratory and its subcontractors. The latter support was carried out through workshops, by documenting the methodology to be used in a set of guides, and through periodic review of the technical activity. The result of this effort to date includes a set of procedure guides, a draft final report on the Level 1 PRA for internal events (excluding internal fires and floods), and progress reports on the fire, flood, and seismic analysis. It is the authors belief that the type of assistance provided by the NRC has been instrumental in assuring a quality product and transferring important technology for use by regulators and operators of Soviet-designed reactors. After a thorough review, the report will be finalized, lessons learned will be applied in the regulatory and operational regimes in the Russian Federation, and consideration will be given to supporting a containment analysis in order to complete a simplified Level 2 PRA

Augmenting Probabilistic Risk Assesment with Malevolent Initiators

International Nuclear Information System (INIS)

Smith, Curtis; Schwieder, David

2011-01-01

As commonly practiced, the use of probabilistic risk assessment (PRA) in nuclear power plants only considers accident initiators such as natural hazards, equipment failures, and human error. Malevolent initiators are ignored in PRA, but are considered the domain of physical security, which uses vulnerability assessment based on an officially specified threat (design basis threat). This paper explores the implications of augmenting and extending existing PRA models by considering new and modified scenarios resulting from malevolent initiators. Teaming the augmented PRA models with conventional vulnerability assessments can cost-effectively enhance security of a nuclear power plant. This methodology is useful for operating plants, as well as in the design of new plants. For the methodology, we have proposed an approach that builds on and extends the practice of PRA for nuclear power plants for security-related issues. Rather than only considering 'random' failures, we demonstrated a framework that is able to represent and model malevolent initiating events and associated plant impacts.

Objective techniques for psychological assessment, phase 2. [techniques for measuring human performance during space flight stress

Science.gov (United States)

Wortz, E. C.; Saur, A. J.; Nowlis, D. P.; Kendall, M. P.

1974-01-01

Results are presented of an initial experiment in a research program designed to develop objective techniques for psychological assessment of individuals and groups participating in long-duration space flights. Specifically examined is the rationale for utilizing measures of attention as an objective assessment technique. Subjects participating in the experiment performed various tasks (eg, playing matrix games which appeared on a display screen along with auditory stimuli). The psychophysiological reactions of the subjects were measured and are given. Previous research of various performance and psychophysiological methods of measuring attention is also discussed. The experiment design (independent and dependent variables) and apparatus (computers and display devices) are described and shown. Conclusions and recommendations are presented.

Risk assessment techniques with applicability in marine engineering

Science.gov (United States)

Rudenko, E.; Panaitescu, F. V.; Panaitescu, M.

2015-11-01

Nowadays risk management is a carefully planned process. The task of risk management is organically woven into the general problem of increasing the efficiency of business. Passive attitude to risk and awareness of its existence are replaced by active management techniques. Risk assessment is one of the most important stages of risk management, since for risk management it is necessary first to analyze and evaluate risk. There are many definitions of this notion but in general case risk assessment refers to the systematic process of identifying the factors and types of risk and their quantitative assessment, i.e. risk analysis methodology combines mutually complementary quantitative and qualitative approaches. Purpose of the work: In this paper we will consider as risk assessment technique Fault Tree analysis (FTA). The objectives are: understand purpose of FTA, understand and apply rules of Boolean algebra, analyse a simple system using FTA, FTA advantages and disadvantages. Research and methodology: The main purpose is to help identify potential causes of system failures before the failures actually occur. We can evaluate the probability of the Top event.The steps of this analize are: the system's examination from Top to Down, the use of symbols to represent events, the use of mathematical tools for critical areas, the use of Fault tree logic diagrams to identify the cause of the Top event. Results: In the finally of study it will be obtained: critical areas, Fault tree logical diagrams and the probability of the Top event. These results can be used for the risk assessment analyses.

Toward risk assessment 2.0: Safety supervisory control and model-based hazard monitoring for risk-informed safety interventions

International Nuclear Information System (INIS)

Favarò, Francesca M.; Saleh, Joseph H.

2016-01-01

Probabilistic Risk Assessment (PRA) is a staple in the engineering risk community, and it has become to some extent synonymous with the entire quantitative risk assessment undertaking. Limitations of PRA continue to occupy researchers, and workarounds are often proposed. After a brief review of this literature, we propose to address some of PRA's limitations by developing a novel framework and analytical tools for model-based system safety, or safety supervisory control, to guide safety interventions and support a dynamic approach to risk assessment and accident prevention. Our work shifts the emphasis from the pervading probabilistic mindset in risk assessment toward the notions of danger indices and hazard temporal contingency. The framework and tools here developed are grounded in Control Theory and make use of the state-space formalism in modeling dynamical systems. We show that the use of state variables enables the definition of metrics for accident escalation, termed hazard levels or danger indices, which measure the “proximity” of the system state to adverse events, and we illustrate the development of such indices. Monitoring of the hazard levels provides diagnostic information to support both on-line and off-line safety interventions. For example, we show how the application of the proposed tools to a rejected takeoff scenario provides new insight to support pilots’ go/no-go decisions. Furthermore, we augment the traditional state-space equations with a hazard equation and use the latter to estimate the times at which critical thresholds for the hazard level are (b)reached. This estimation process provides important prognostic information and produces a proxy for a time-to-accident metric or advance notice for an impending adverse event. The ability to estimate these two hazard coordinates, danger index and time-to-accident, offers many possibilities for informing system control strategies and improving accident prevention and risk mitigation

MATILDA Version-2: Rough Earth TIALD Model for Laser Probabilistic Risk Assessment in Hilly Terrain - Part II

Science.gov (United States)

2017-07-28

risk assessment for “unsafe” scenarios. Recently, attention in the DoD has turned to Probabilistic Risk Assessment (PRA) models [5,6] as an...corresponding to the CRA undershoot boundary. The magenta- coloured line represents the portion of the C-RX(U) circle that would contribute to the...Tertiary Precaution Surface. Undershoot related laser firing restrictions within the green- coloured C-RX(U) can be ignored. Figure 34

Risk Management at NASA and Its Applicability to the Oil and Gas Industry

Science.gov (United States)

Kaplan, David

2018-01-01

NASA has a world-class capability for quantitatively assessing the risk of highly-complex, isolated engineering structures operated in extremely hostile environments. In particular, the International Space Station (ISS) represents a reasonable risk analog for High Pressure, High Temperature drilling and production operations on deepwater rigs. Through a long-term U.S. Government Interagency Agreement, BSEE has partnered with NASA to modify NASA's Probabilistic Risk Assessment (PRA) capabilities for application to deepwater drilling and production operations. The immediate focus of the activity will be to modify NASA PRA Procedure Guides and Methodology Documents to make them applicable to the Oil &Gas Industry. The next step will be for NASA to produce a PRA for a critical drilling system component, such as a Blowout Preventer (BOP). Subsequent activities will be for NASA and industry partners to jointly develop increasingly complex PRA's that analyze other critical drilling and production system components, including both hardware and human reliability. In the presentation, NASA will provide the objectives, schedule, and current status of its PRA activities for BSEE. Additionally, NASA has a Space Act Agreement with Anadarko Petroleum Corporation to develop a PRA for a generic 20K BOP. NASA will summarize some of the preliminary insights gained to date from that 20K BOP PRA as an example of the distinction between quantitative versus qualitative risk assessment.

External event Probabilistic Risk Assessment for the High Flux Isotope Reactor (HFIR)

International Nuclear Information System (INIS)

Flanagan, G.F.; Johnson, D.H.; Buttemer, D.; Perla, H.F.; Chien, S.H.

1989-01-01

The High Flux Isotope Reactor (HFIR) is a high performance isotope production and research reactor which has been in operation at Oak Ridge National Laboratory (ORNL) since 1965. In late 1986 the reactor was shut down as a result of discovery of unexpected neutron embrittlement of the reactor vessel. In January of 1988 a level 1 Probabilistic Risk Assessment (PRA) (excluding external events) was published as part of the response to the many reviews that followed the shutdown and for use by ORNL to prioritize action items intended to upgrade the safety of the reactor. A conservative estimate of the core damage frequency initiated by internal events for HFIR was 3.11 x 10 -4 . In June 1989 a draft external events initiated PRA was published. The dominant contributions from external events came from seismic, wind, and fires. The overall external event contribution to core damage frequency is about 50% of the internal event initiated contribution and is dominated by seismic events

Probabilistic risk assessment methodology

International Nuclear Information System (INIS)

Shinaishin, M.A.

1988-06-01

The objective of this work is to provide the tools necessary for clear identification of: the purpose of a Probabilistic Risk Study, the bounds and depth of the study, the proper modeling techniques to be used, the failure modes contributing to the analysis, the classical and baysian approaches for manipulating data necessary for quantification, ways for treating uncertainties, and available computer codes that may be used in performing such probabilistic analysis. In addition, it provides the means for measuring the importance of a safety feature to maintaining a level of risk at a Nuclear Power Plant and the worth of optimizing a safety system in risk reduction. In applying these techniques so that they accommodate our national resources and needs it was felt that emphasis should be put on the system reliability analysis level of PRA. Objectives of such studies could include: comparing systems' designs of the various vendors in the bedding stage, and performing grid reliability and human performance analysis using national specific data. (author)

Probabilistic risk assessment methodology

Energy Technology Data Exchange (ETDEWEB)

Shinaishin, M A

1988-06-15

The objective of this work is to provide the tools necessary for clear identification of: the purpose of a Probabilistic Risk Study, the bounds and depth of the study, the proper modeling techniques to be used, the failure modes contributing to the analysis, the classical and baysian approaches for manipulating data necessary for quantification, ways for treating uncertainties, and available computer codes that may be used in performing such probabilistic analysis. In addition, it provides the means for measuring the importance of a safety feature to maintaining a level of risk at a Nuclear Power Plant and the worth of optimizing a safety system in risk reduction. In applying these techniques so that they accommodate our national resources and needs it was felt that emphasis should be put on the system reliability analysis level of PRA. Objectives of such studies could include: comparing systems' designs of the various vendors in the bedding stage, and performing grid reliability and human performance analysis using national specific data. (author)

A social impact assessment of the floodwater spreading project on the Gareh-Bygone plain in Iran: A causal comparative approach

International Nuclear Information System (INIS)

Ahmadvand, Mostafa; Karami, Ezatollah

2009-01-01

The purpose of this study was to explore the social impacts of the floodwater spreading project (FWSP) on the Gareh-Bygone plain, Iran. The study was in the form of a causal comparative design, and a triangulation technique was used to collect data including the use of survey data, archival data, and a participatory rural appraisal (PRA). The causal comparative method requires a comparison of villages with and without the FWSP. Therefore, a survey was conducted using stratified random sampling to select 202 households in villages with and without FWSP in the plain. Significant differences were found between the respondents in villages with and without FWSP with regard to social impact criteria. In spite of the project had negative impact on perceived wellbeing, social capital, social structure development; it had positive impact on quality of life, rural and agricultural economic conditions, and conservation of community resources. However, no significant difference was found between women and men regarding the SIA of FWSP in Gareh-Bygone plain. Analysis of the archival data and PRA techniques supported the survey results and demonstrated that the project improved environmental criteria and deteriorated social dimensions

Prioritization of motor operated valves based on risk importances

International Nuclear Information System (INIS)

Vesely, W.E.; Weidenhamer, G.H.

1994-01-01

The plant Probabilistic Risk Assessment (PRA) can be a potentially useful and powerful tool for helping to define an effective response to GL 89-10. The plant PRA can be used to prioritize the Motor Operated Valves (MOV) dynamic test. The plant PRA can also be used to determine test schedules for the MOVs. In order for the PRA to be validly used to respond to GL 89-10, various issues need to be validly addressed. Eleven issues are specifically identified and responses to these issues are outlined. The issues of joint MOV importance, PRA truncation, and validation of the proposed approach are specifically highlighted and more detailed response considerations are described. As in all PRA applications, sensitivity studies and uncertainty considerations should be incorporated in the PRA evaluations. 4 refs, 3 tabs

Presença de ovos de Toxocara spp em praças públicas da cidade de Uberlândia, Minas Gerais, Brasil Presence of Toxocara spp eggs in public squares of Uberlândia city, Minas Gerais, Brazil

Directory of Open Access Journals (Sweden)

Julia Maria Costa-Cruz

1994-02-01

Full Text Available O objetivo deste trabalho foi verificar a presença de ovos de Toxocara spp nos solos de praças públicas da cidade de Uberlândia, região do Triângulo Mineiro, no período de outubro de 1991 a janeiro de 1992. A cidade possui 89 praças distribuídas em 39 bairros. Para que se tivesse um perfil da ocorrência do parasita, foram colhidas amostras de terra e areia (quando existente de uma praça, determinada por sorteio, por bairro. As amostras de solo homogeneizadas de 5 pontos distintos das 39 praças foram colhidas em frascos plásticos e analisadas empregando-se os métodos de flutuação em solução saturada de cloreto de sódio e de solução saturada de sulfato de magnésio contendo 5% de iodeto de potássio. Para cada amostra foram realizados os dois métodos em duplicata. Os resultados mostraram estarem contaminados os solos de 9 praças (23,07%, das quais 6 localizavam-se próximas ao centro da cidade. A realização de mais de um método laboratorial permitiu a identificação do agente em maior porcentagem de locais.The objective of this study was to verify the presence of Toxocara spp eggs in the public squares ground in Uberlândia, region of Triângulo Mineiro, from October 1991 to January 1992. The city has 89 squares distributed in 39 boroughs. In order to have an outline of the parasite's occurrence, ground and sand (when existent samples were collected from one square per boroughors. The squares were determined by draw. The homogenized ground samples of distinct points of the 39 squares, were colected in plastic flasks and analysed by the methods of fluctuation in sodium chloride saturated solution and magnesium sulphate saturated solution which contained 5% of potassium iodate. Both methods were executed twice for each sample. The results showed that the ground of 9 squares (23.07% was contaminated, 6 of which were located near downtown. The execution of more than one laboratorial method permited the agent's identification

A probabilistic risk assessment for field radiography based on expert judgment and opinion

International Nuclear Information System (INIS)

Jang, Han-Ki; Ryu, Hyung-Joon; Kim, Ji-Young; Lee, Jai-Ki; Cho, Kun-Woo

2011-01-01

A probabilistic approach was applied to assess radiation risk associated with the field radiography using gamma sources. The Delphi method based on the expert judgments and opinions was used in the process of characterization of parameters affecting risk, which are inevitably subject to large uncertainties. A mathematical approach applying the Bayesian inferences was employed for data processing to improve the Delphi results. This process consists of three phases: (1) setting prior distributions, (2) constructing the likelihood functions and (3) deriving the posterior distributions based on the likelihood functions. The approach for characterizing input parameters using the Bayesian inference is provided for improved risk estimates without intentional rejection of part of the data, which demonstrated utility of Bayesian updating of distributions of uncertain input parameters in PRA (Probabilistic Risk Assessment). The data analysis portion for PRA in field radiography is addressed for estimates of the parameters used to determine the frequencies and consequences of the various events modeled. In this study, radiological risks for the worker and the public member in the vicinity of the work place are estimated for field radiography system in Korea based on two-dimensional Monte Carlo Analysis (2D MCA). (author)

Review of seismic probabilistic risk assessment and the use of sensitivity analysis

International Nuclear Information System (INIS)

Shiu, K.K.; Reed, J.W.; McCann, M.W. Jr.

1985-01-01

This paper presents results of sensitivity reviews performed to address a range of questions which arise in the context of seismic probabilistic risk assessment (PRA). In a seismic PRA, sensitivity evaluations can be divided into three areas: hazard, fragility, and system modeling. As a part of the review of standard boiling water reactor seismic PRA which was performed by General Electric (GE), a reassessment of the plant damage states frequency and a detailed sensitivity analysis were conducted at Brookhaven National Laboratory. The rationale for such an undertaking is that in this case: (1) the standard plant may be sited anywhere in the eastern US (i.e., in regions with safety shutdown earthquake (SSE) values equal to or less than 0.3g peak ground acceleration), (2) it may have equipment whose fragility values could vary over a wide range; and (3) there are variations in system designs outside the original defined scope. Seismic event trees and fault trees were developed to model the different system and plant accident sequences. Hazard curves which represent various sites on the east coast were obtained; alternate structure and equipment fragility data were postulated. Various combinations of hazard and fragility data were analyzed. In addition, system modeling was perturbed to examine the impact upon the final results. Orders of magnitude variation were observed in the plant damage state frequency among the different cases. 7 references, 3 figures, 2 tables

Assessment of attention and different feedback techniques on ...

African Journals Online (AJOL)

Assessment of attention and different feedback techniques on psychomotor skill learninb and retention. JO Adeogun, IA Oyewuni. Abstract. No Abstract. Full Text: EMAIL FULL TEXT EMAIL FULL TEXT · DOWNLOAD FULL TEXT DOWNLOAD FULL TEXT · http://dx.doi.org/10.4314/ajpherd.v11i4.46288 · AJOL African ...

Praça da Matemática: as faces da história na construção de um monumento

OpenAIRE

Pimentel, Augusto Cesar Aguiar

2008-01-01

O presente trabalho consiste na investigação histórico-cultural da construção de uma praça, com um monumento em homenagem à Matemática, idealizado por um prefeito visionário, de 29 anos, na cidade de Itaocara, interior do Estado do Rio de Janeiro, no ano de 1943. O que é hoje uma atração turística, surgiu da criatividade do jovem político que apesar de advogado, via na matemática sua fonte de prazer. Para essa pesquisa, realizou-se entrevistas com o próprio prefeito da época...

Combining heuristic and statistical techniques in landslide hazard assessments

Science.gov (United States)

Cepeda, Jose; Schwendtner, Barbara; Quan, Byron; Nadim, Farrokh; Diaz, Manuel; Molina, Giovanni

2014-05-01

As a contribution to the Global Assessment Report 2013 - GAR2013, coordinated by the United Nations International Strategy for Disaster Reduction - UNISDR, a drill-down exercise for landslide hazard assessment was carried out by entering the results of both heuristic and statistical techniques into a new but simple combination rule. The data available for this evaluation included landslide inventories, both historical and event-based. In addition to the application of a heuristic method used in the previous editions of GAR, the availability of inventories motivated the use of statistical methods. The heuristic technique is largely based on the Mora & Vahrson method, which estimates hazard as the product of susceptibility and triggering factors, where classes are weighted based on expert judgment and experience. Two statistical methods were also applied: the landslide index method, which estimates weights of the classes for the susceptibility and triggering factors based on the evidence provided by the density of landslides in each class of the factors; and the weights of evidence method, which extends the previous technique to include both positive and negative evidence of landslide occurrence in the estimation of weights for the classes. One key aspect during the hazard evaluation was the decision on the methodology to be chosen for the final assessment. Instead of opting for a single methodology, it was decided to combine the results of the three implemented techniques using a combination rule based on a normalization of the results of each method. The hazard evaluation was performed for both earthquake- and rainfall-induced landslides. The country chosen for the drill-down exercise was El Salvador. The results indicate that highest hazard levels are concentrated along the central volcanic chain and at the centre of the northern mountains.

Application of Isotope Techniques for Assessing Nutrient Dynamics in River Basins

International Nuclear Information System (INIS)

2013-05-01

Nutrients are necessary for the growth and survival of animals, plants and other organisms. However, industrial, agricultural and urban development has dramatically increased nutrient levels in river systems, including nitrogen and phosphorus containing substances, degrading water quality, causing acidification and eutrophication and affecting aquatic ecosystems. Nutrient assessment and management in river systems has been an important part of water resource management for the past few decades, but the provision of appropriate and effective nutrient assessment and management continues to be a challenge for water resource managers and policy makers. Difficulties in assessment and management are due in part to the fact that nutrients in rivers may originate from a variety of sources, take numerous pathways and transform into other substances. This publication presents the application of isotope techniques as a powerful tool for evaluating the sources, pathways, transformation, and fate of nutrients in river systems, focusing on nitrogen, phosphorus and carbon containing substances. Eleven researchers using various isotope techniques for different aspects of nutrient studies and two IAEA officers met in a technical meeting and discussed a publication that could assist water resource managers in dealing with nutrient assessment and management issues in river systems. These researchers also recognized the need for careful consideration in selecting appropriate isotope techniques in view of not only technical, but also financial, human resources and logistical capabilities, among others. These contributors are listed as major authors in the later pages of this document. This publication aims at serving water resource managers as a guidebook on the application of isotope techniques in nutrient assessment and management, but it is also expected to be of practical aid for other interested and concerned individuals and organization.

Application of Isotope Techniques for Assessing Nutrient Dynamics in River Basins

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-05-15

Nutrients are necessary for the growth and survival of animals, plants and other organisms. However, industrial, agricultural and urban development has dramatically increased nutrient levels in river systems, including nitrogen and phosphorus containing substances, degrading water quality, causing acidification and eutrophication and affecting aquatic ecosystems. Nutrient assessment and management in river systems has been an important part of water resource management for the past few decades, but the provision of appropriate and effective nutrient assessment and management continues to be a challenge for water resource managers and policy makers. Difficulties in assessment and management are due in part to the fact that nutrients in rivers may originate from a variety of sources, take numerous pathways and transform into other substances. This publication presents the application of isotope techniques as a powerful tool for evaluating the sources, pathways, transformation, and fate of nutrients in river systems, focusing on nitrogen, phosphorus and carbon containing substances. Eleven researchers using various isotope techniques for different aspects of nutrient studies and two IAEA officers met in a technical meeting and discussed a publication that could assist water resource managers in dealing with nutrient assessment and management issues in river systems. These researchers also recognized the need for careful consideration in selecting appropriate isotope techniques in view of not only technical, but also financial, human resources and logistical capabilities, among others. These contributors are listed as major authors in the later pages of this document. This publication aims at serving water resource managers as a guidebook on the application of isotope techniques in nutrient assessment and management, but it is also expected to be of practical aid for other interested and concerned individuals and organization.

Establishment of safety goal and its quantification based on risk assessment

International Nuclear Information System (INIS)

Miyano, Hiroshi; Muramatsu, Ken

2017-01-01

We must clarify the safety objectives sought by society in securing the safety of nuclear reactors and nuclear power plants. For that purpose, it is useful to utilize risk assessment. Quantitative methods including probabilistic risk assessment (PRA) are superior in terms of scientific rationality and quantitative performance compared with conventional deterministic methods, and able to indicate an objective numerical value of safety level. Consequently, quantitative methods can enhance the transparency, consistency, compliance, predictability, and explanatory power of regulatory decisions toward business operators and citizens. Business operators can explain the validity of their own safety assurance activities to regulators and citizens. The goal to be secured becomes clear by incorporating the safety goal into the specific performance goal required for the nuclear power plant from the viewpoint of deep safeguard, and it becomes easy to evaluate the effectiveness of the safety measures. It helps us greatly in judging and selecting the appropriateness of safety measures. It should be noted: the fact that the result of implementing the PRA satisfies the safety goal is not a sufficient condition in the sense of guaranteeing complete safety but a necessary condition. The nuclear power field is a region with large uncertainty, and research/efforts for accuracy improvement and evaluation validity will be required continuously. (A.O.)

Operational Performance Risk Assessment in Support of A Supervisory Control System

Energy Technology Data Exchange (ETDEWEB)

Denning, Richard S. [Self Employed; Muhlheim, Michael David [ORNL; Cetiner, Sacit M. [ORNL; Guler Yigitoglu, Askin [ORNL

2017-06-01

Supervisory control system (SCS) is developed for multi-unit advanced small modular reactors to minimize human interventions in both normal and abnormal operations. In SCS, control action decisions made based on probabilistic risk assessment approach via Event Trees/Fault Trees. Although traditional PRA tools are implemented, their scope is extended to normal operations and application is reversed; success of non-safety related system instead failure of safety systems this extended PRA approach called as operational performance risk assessment (OPRA). OPRA helps to identify success paths, combination of control actions for transients and to quantify these success paths to provide possible actions without activating plant protection system. In this paper, a case study of the OPRA in supervisory control system is demonstrated within the context of the ALMR PRISM design, specifically power conversion system. The scenario investigated involved a condition that the feed water control valve is observed to be drifting to the closed position. Alternative plant configurations were identified via OPRA that would allow the plant to continue to operate at full or reduced power. Dynamic analyses were performed with a thermal-hydraulic model of the ALMR PRISM system using Modelica to evaluate remained safety margins. Successful recovery paths for the selected scenario are identified and quantified via SCS.

Assessment of heterogeneous geological environment using geostatistical techniques

International Nuclear Information System (INIS)

Toida, Masaru; Suyama, Yasuhiro; Shiogama, Yukihiro; Atsumi, Hiroyuki; Abe, Yasunori; Furuichi, Mitsuaki

2003-02-01

'Geoscientific' research at Tono are developing site investigation and assessment techniques in geological environment. One of their important themes is to establish rational methodology to reduce uncertainties associated with the understanding of geological environment, which often exhibits significant heterogeneity. Purpose of this study is to identify and evaluate uncertainties associated with the understanding of geological environment. Because it is useful to guide designing effective site investigation techniques to reduce the uncertainty. For this, a methodology of the uncertainty analysis concerning the heterogeneous geological environment has been developed. In this report the methodology has also been tested through an exercise attempted in Tono area to demonstrate its applicability. This report summarizes as follows: 1) The exercise shows that the methodology considered 'variability' and 'ignorance' can demonstrate its applicability at three-dimensional case. 2) The exercise shows that the methodology can identity and evaluate uncertainties concerning ground water flow associated with performance assessment. 3) Based on sensitivity analyses, it is possible for the methodology to support designs of the following stage investigations to reduce the uncertainties efficiently. (author)

[Research progress and development trend of quantitative assessment techniques for urban thermal environment.

Science.gov (United States)

Sun, Tie Gang; Xiao, Rong Bo; Cai, Yun Nan; Wang, Yao Wu; Wu, Chang Guang

2016-08-01

Quantitative assessment of urban thermal environment has become a focus for urban climate and environmental science since the concept of urban heat island has been proposed. With the continual development of space information and computer simulation technology, substantial progresses have been made on quantitative assessment techniques and methods of urban thermal environment. The quantitative assessment techniques have been developed to dynamics simulation and forecast of thermal environment at various scales based on statistical analysis of thermal environment on urban-scale using the historical data of weather stations. This study reviewed the development progress of ground meteorological observation, thermal infrared remote sensing and numerical simulation. Moreover, the potential advantages and disadvantages, applicability and the development trends of these techniques were also summarized, aiming to add fundamental knowledge of understanding the urban thermal environment assessment and optimization.

The in-depth safety assessment (ISA) pilot projects in Ukraine

International Nuclear Information System (INIS)

Kot, C. A.

1998-01-01

Ukraine operates pressurized water reactors of the Soviet-designed type, VVER. All Ukrainian plants are currently operating with annually renewable permits until they update their safety analysis reports (SARs). After approval of the SARS by the Ukrainian Nuclear Regulatory Authority, the plants will be granted longer-term operating licenses. In September 1995, the Nuclear Regulatory Authority and the Government Nuclear Power Coordinating Committee of Ukraine issued a new contents requirement for the safety analysis reports of VVERs in Ukraine. It contains requirements in three major areas: design basis accident (DBA) analysis, probabilistic risk assessment (PRA), and beyond design-basis accident (BDBA) analysis. The DBA requirements are an expanded version of the older SAR requirements. The last two requirements, on PRA and BDBA, are new. The US Department of Energy (USDOE), through the International Nuclear Safety Program (INSP), has initiated an assistance and technology transfer program to Ukraine to assist their nuclear power stations in developing a Western-type technical basis for the new SARS. USDOE sponsored In-Depth Safety Assessments (ISAs) have been initiated at three pilot nuclear reactor units in Ukraine, South Ukraine Unit 1, Zaporizhzhya Unit 5, and Rivne Unit 1. USDOE/INSP have structured the ISA program in such a way as to provide maximum assistance and technology transfer to Ukraine while encouraging and supporting the Ukrainian plants to take the responsibility and initiative and to perform the required assessments

Modeling and Quantification of Team Performance in Human Reliability Analysis for Probabilistic Risk Assessment

Energy Technology Data Exchange (ETDEWEB)

Jeffrey C. JOe; Ronald L. Boring

2014-06-01

Probabilistic Risk Assessment (PRA) and Human Reliability Assessment (HRA) are important technical contributors to the United States (U.S.) Nuclear Regulatory Commission’s (NRC) risk-informed and performance based approach to regulating U.S. commercial nuclear activities. Furthermore, all currently operating commercial NPPs in the U.S. are required by federal regulation to be staffed with crews of operators. Yet, aspects of team performance are underspecified in most HRA methods that are widely used in the nuclear industry. There are a variety of "emergent" team cognition and teamwork errors (e.g., communication errors) that are 1) distinct from individual human errors, and 2) important to understand from a PRA perspective. The lack of robust models or quantification of team performance is an issue that affects the accuracy and validity of HRA methods and models, leading to significant uncertainty in estimating HEPs. This paper describes research that has the objective to model and quantify team dynamics and teamwork within NPP control room crews for risk informed applications, thereby improving the technical basis of HRA, which improves the risk-informed approach the NRC uses to regulate the U.S. commercial nuclear industry.

A probabilistic risk assessment of the LLNL Plutonium facility's evaluation basis fire operational accident

International Nuclear Information System (INIS)

Brumburgh, G.

1994-01-01

The Lawrence Livermore National Laboratory (LLNL) Plutonium Facility conducts numerous involving plutonium to include device fabrication, development of fabrication techniques, metallurgy research, and laser isotope separation. A Safety Analysis Report (SAR) for the building 332 Plutonium Facility was completed rational safety and acceptable risk to employees, the public, government property, and the environment. This paper outlines the PRA analysis of the Evaluation Basis Fire (EDF) operational accident. The EBF postulates the worst-case programmatic impact event for the Plutonium Facility

Coping with uncertainty in environmental impact assessments: Open techniques

NARCIS (Netherlands)

Chivatá Cárdenas, Ibsen; Halman, Johannes I.M.

2016-01-01

Uncertainty is virtually unavoidable in environmental impact assessments (EIAs). From the literature related to treating and managing uncertainty, we have identified specific techniques for coping with uncertainty in EIAs. Here, we have focused on basic steps in the decision-making process that take

A Methodology for Modeling Nuclear Power Plant Passive Component Aging in Probabilistic Risk Assessment under the Impact of Operating Conditions, Surveillance and Maintenance Activities

Science.gov (United States)

Guler Yigitoglu, Askin

In the context of long operation of nuclear power plants (NPPs) (i.e., 60-80 years, and beyond), investigation of the aging of passive systems, structures and components (SSCs) is important to assess safety margins and to decide on reactor life extension as indicated within the U.S. Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) Program. In the traditional probabilistic risk assessment (PRA) methodology, evaluating the potential significance of aging of passive SSCs on plant risk is challenging. Although passive SSC failure rates can be added as initiating event frequencies or basic event failure rates in the traditional event-tree/fault-tree methodology, these failure rates are generally based on generic plant failure data which means that the true state of a specific plant is not reflected in a realistic manner on aging effects. Dynamic PRA methodologies have gained attention recently due to their capability to account for the plant state and thus address the difficulties in the traditional PRA modeling of aging effects of passive components using physics-based models (and also in the modeling of digital instrumentation and control systems). Physics-based models can capture the impact of complex aging processes (e.g., fatigue, stress corrosion cracking, flow-accelerated corrosion, etc.) on SSCs and can be utilized to estimate passive SSC failure rates using realistic NPP data from reactor simulation, as well as considering effects of surveillance and maintenance activities. The objectives of this dissertation are twofold: The development of a methodology for the incorporation of aging modeling of passive SSC into a reactor simulation environment to provide a framework for evaluation of their risk contribution in both the dynamic and traditional PRA; and the demonstration of the methodology through its application to pressurizer surge line pipe weld and steam generator tubes in commercial nuclear power plants. In the proposed methodology, a

Techniques for Liquid Rocket Combustion Spontaneous Stability and Rough Combustion Assessments

Science.gov (United States)

Kenny, R. J.; Giacomoni, C.; Casiano, M. J.; Fischbach, S. R.

2016-01-01

This work presents techniques for liquid rocket engine combustion stability assessments with respect to spontaneous stability and rough combustion. Techniques covering empirical parameter extraction, which were established in prior works, are applied for three additional programs: the F-1 Gas Generator (F1GG) component test program, the RS-84 preburner component test program, and the Marshall Integrated Test Rig (MITR) program. Stability assessment parameters from these programs are compared against prior established spontaneous stability metrics and updates are identified. Also, a procedure for comparing measured with predicted mode shapes is presented, based on an extension of the Modal Assurance Criterion (MAC).

Hubungan antara Pengetahuan, Sikap, dan Perilaku terhadap Pemeliharaan Kebersihan Gigi dan Mulut dengan Status Kesehatan Periodontal Pra Lansia di Posbindu Kecamatan Indihiang Kota Tasikmalaya

Directory of Open Access Journals (Sweden)

Culia Rahayu

2014-06-01

Full Text Available Proses penuaan merupakan salah satu faktor sistemik yang mempengaruhi respon tubuh terhadap terjadinya penyakit periodontal. Penelitian ini bertujuan untuk mengetahui hubungan antara pengetahuan, sikap dan perilaku terhadap pemeliharaan kebersihan gigi dan mulut dengan status kesehatan periodontal pra lansia. Penelitian ini merupakan penelitian non eksperimental, dengan rancangan cross sectional. Subjek penelitian berjumlah 225 orang pra lansia dari 9 Posbindu di Kecamatan Indihiang Tasikmalaya, yang diambil dengan cara purposive sampling. Variabel pengaruh terdiri dari pengetahuan, sikap dan perilaku terhadap pemeliharaan kebersihan gigi dan mulut. Variabel pengetahuan diukur menggunakan kuesioner tertutup dengan pilihan jawaban benar atau salah. Variabel sikap dan perilaku diukur menggunakan kuesioner dibuat menurut skala Likert. Kuesioner telah diuji validitas (koefisien korelasi ≥ 0,30 dan uji reliabilitas (alpha cronbach >0,60. Variabel terpengaruh adalah status kesehatan periodontal yang diukur menggunakan indeks CPITN. Analisa data menggunakan analisis korelasi dan regresi berganda. Hasil analisis regresi berganda menunjukkan bahwa variabel pengetahuan, sikap dan perilaku terhadap pemeliharaan kebersihan gigi dan mulut mempunyai hubungan signifikan terhadap status kesehatan periodontal (F =30,681 dan p =0,001, dan memberikan kontribusi pengaruh sebesar 29,4% (R² = 0,294 terhadap status kesehatan periodontal. Perilaku terhadap pemeliharaan kebersihan gigi dan mulut memberikan kontribusi pengaruh paling besar terhadap status kesehatan periodontal yaitu sebesar 6,9%. Kesimpulan dari penelitian ini adalah semakin baik pengetahuan, sikap dan perilaku terhadap pemeliharaan kebersihan gigi dan mulut, semakin baik status kesehatan periodontal pra lansia. Perilaku terhadap pemeliharaan kebersihan gigi dan mulut memberikan kontribusi pengaruh paling besar terhadap status kesehatan periodontal pra lansia. Correlation Between Knowledge, Attitude

Development of infrastructure for the regulatory authority to implement risk-informed regulation

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

It is important to assure the technical adequacy of probabilistic risk assessment (PRA) to implement risk-informed regulation of nuclear power plants (NPPs). JNES has been conducting various activities, such as development of PRA model, method, and data base, in order to assure the technical adequacy of PRA as development of the infrastructure for the regulatory authority to implement risk-informed regulation. In 2012, JNES updated the reliability data base used in PRA and improved PRA models to enhance the technical bases of PRA. In addition, JNES has been establishing the PRA model for fuel damage in the spent fuel storage pool in NPPs. As for improvement of PRA model for core damage in reactor, JNES conducted the study including feasibility of a simplified reliability model for digital I and C system developed by the digital I and C task group of OECD/NEA CSNI WGRISK by reproducing the sample calculation, and improvement of PRA models of individual NPPs in Japan. JNES is making effort to develop the procedures of internal fire PRA and internal flooding PRA. To improve the internal fire PRA, JNES is participating in OECD/NEA FIRE project to obtain the latest information and to validate and improve the fire propagation analysis codes and the parameters. JNES is establishing a method for analyzing internal influence due to flooding in NPPs, and this method is the base to develop the procedure of internal flooding PRA. (author)

Review of the Shoreham Nuclear Power Station Probabilistic Risk Assessment: internal events and core damage frequency

International Nuclear Information System (INIS)

Ilberg, D.; Shiu, K.; Hanan, N.; Anavim, E.

1985-11-01

A review of the Probabilistic Risk Assessment of the Shoreham Nuclear Power Station was conducted with the broad objective of evaluating its risks in relation to those identified in the Reactor Safety Study (WASH-1400). The scope of the review was limited to the ''front end'' part, i.e., to the evaluation of the frequencies of states in which core damage may occur. Furthermore, the review considered only internally generated accidents, consistent with the scope of the PRA. The review included an assessment of the assumptions and methods used in the Shoreham study. It also encompassed a reevaluation of the main results within the scope and general methodological framework of the Shoreham PRA, including both qualitative and quantitative analyses of accident initiators, data bases, and accident sequences which result in initiation of core damage. Specific comparisons are given between the Shoreham study, the results of the present review, and the WASH-1400 BWR, for the core damage frequency. The effect of modeling uncertainties was considered by a limited sensitivity study so as to show how the results would change if other assumptions were made. This review provides an independently assessed point value estimate of core damage frequency and describes the major contributors, by frontline systems and by accident sequences. 17 figs., 81 tabs

Preliminary risk assessment of the Integral Inherently-Safe Light Water Reactor

International Nuclear Information System (INIS)

McCarroll, Kellen R.; Lee, John C.; Manera, Annalisa; Memmott, Matthew J.; Ferroni, Paolo

2017-01-01

The Integral, Inherently Safe Light Water Reactor (I 2 S-LWR) concept seeks to significantly increase nuclear power plant safety. The project implements a safety-by-design philosophy, eliminating several initiating events and providing novel, passive safety systems at the conceptual phase. Pursuit of unparalleled safety employs an integrated development process linking design with deterministic and probabilistic safety analyses. Unique aspects of the I 2 S-LWR concept and design process present challenges to the probabilistic risk assessment (PRA), particularly regarding overall flexibility, auditability and resolution of results. Useful approaches to initiating events and conditional failures are presented. To exemplify the risk-informed design process using PRA, a trade-off study of two safety system configurations is presented. Although further optimization is required, preliminary results indicate that the I 2 S-LWR can achieve a core damage frequency (CDF) from internal events less than 1.01 × 10 −8 /ry, including reactor vessel ruptures. Containment bypass frequency due to primary heat exchanger rupture is found to be comparable to non-vessel rupture CDF.

Dynamical systems probabilistic risk assessment

Energy Technology Data Exchange (ETDEWEB)

Denman, Matthew R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Ames, Arlo Leroy [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2014-03-01

Probabilistic Risk Assessment (PRA) is the primary tool used to risk-inform nuclear power regulatory and licensing activities. Risk-informed regulations are intended to reduce inherent conservatism in regulatory metrics (e.g., allowable operating conditions and technical specifications) which are built into the regulatory framework by quantifying both the total risk profile as well as the change in the risk profile caused by an event or action (e.g., in-service inspection procedures or power uprates). Dynamical Systems (DS) analysis has been used to understand unintended time-dependent feedbacks in both industrial and organizational settings. In dynamical systems analysis, feedback loops can be characterized and studied as a function of time to describe the changes to the reliability of plant Structures, Systems and Components (SSCs). While DS has been used in many subject areas, some even within the PRA community, it has not been applied toward creating long-time horizon, dynamic PRAs (with time scales ranging between days and decades depending upon the analysis). Understanding slowly developing dynamic effects, such as wear-out, on SSC reliabilities may be instrumental in ensuring a safely and reliably operating nuclear fleet. Improving the estimation of a plant's continuously changing risk profile will allow for more meaningful risk insights, greater stakeholder confidence in risk insights, and increased operational flexibility.

Tortura e transição democrática: uma análise histórica do filme “Pra frente, Brasil”

Directory of Open Access Journals (Sweden)

Paulo Jorge Corrêa Campos

2014-06-01

Full Text Available O presente trabalho pretende analisar o filme “Pra frente, Brasil” situando o mesmo no contexto da abertura democrática dos anos 1980. Para tal, busca-se tecer o contexto e o lugar social onde foi concebido o filme. Na análise, atenta-se para o papel da tortura presente na obra cinematográfica e para os elementos semióticos usados no filme para incutir uma visão específica no espectador. Ademais, considerando-se que o mesmo sofre censura prévia do governo, procura-se questionar a superficialidade da democracia mesmo no agonizar da ditadura civil-militar.

The Importance of HRA in Human Space Flight: Understanding the Risks

Science.gov (United States)

Hamlin, Teri

2010-01-01

Human performance is critical to crew safety during space missions. Humans interact with hardware and software during ground processing, normal flight, and in response to events. Human interactions with hardware and software can cause Loss of Crew and/or Vehicle (LOCV) through improper actions, or may prevent LOCV through recovery and control actions. Humans have the ability to deal with complex situations and system interactions beyond the capability of machines. Human Reliability Analysis (HRA) is a method used to qualitatively and quantitatively assess the occurrence of human failures that affect availability and reliability of complex systems. Modeling human actions with their corresponding failure probabilities in a Probabilistic Risk Assessment (PRA) provides a more complete picture of system risks and risk contributions. A high-quality HRA can provide valuable information on potential areas for improvement, including training, procedures, human interfaces design, and the need for automation. Modeling human error has always been a challenge in part because performance data is not always readily available. For spaceflight, the challenge is amplified not only because of the small number of participants and limited amount of performance data available, but also due to the lack of definition of the unique factors influencing human performance in space. These factors, called performance shaping factors in HRA terminology, are used in HRA techniques to modify basic human error probabilities in order to capture the context of an analyzed task. Many of the human error modeling techniques were developed within the context of nuclear power plants and therefore the methodologies do not address spaceflight factors such as the effects of microgravity and longer duration missions. This presentation will describe the types of human error risks which have shown up as risk drivers in the Shuttle PRA which may be applicable to commercial space flight. As with other large PRAs

The EBR-II Probabilistic Risk Assessment: Results and insights

International Nuclear Information System (INIS)

Hill, D.J.; Ragland, W.A.; Roglans, J.

1993-01-01

This paper summarizes the results from the recently completed EBR-II Probabilistic Risk Assessment (PRA) and provides an analysis of the source of risk of the operation of EBR-II from both internal and external initiating events. The EBR-II PRA explicitly accounts for the role of reactivity feedbacks in reducing fuel damage. The results show that the expected core damage frequency from internal initiating events at EBR-II is very low, 1. 6 10 -6 yr -1 , even with a wide definition of core damage (essentially that of exceeding Technical Specification limits). The probability of damage, primarily due to liquid metal fires, from externally initiated events (excluding earthquake) is 3.6 10 -6 yr -1 . overall these results are considerably better than results for other research reactors and the nuclear industry in general and stem from three main sources: low likelihood of loss of coolant due to low system pressure and top entry double, vessels; low likelihood of loss of decay heat removal due to reliance on passive means; and low likelihood of power/flow mismatch due to both passive feedbacks and reliability of rod scram capability

A novel optical investigation technique for railroad track inspection and assessment

Science.gov (United States)

Sabato, Alessandro; Beale, Christopher H.; Niezrecki, Christopher

2017-04-01

Track failures due to cross tie degradation or loss in ballast support may result in a number of problems ranging from simple service interruptions to derailments. Structural Health Monitoring (SHM) of railway track is important for safety reasons and to reduce downtime and maintenance costs. For this reason, novel and cost-effective track inspection technologies for assessing tracks' health are currently insufficient and needed. Advancements achieved in recent years in cameras technology, optical sensors, and image-processing algorithms have made machine vision, Structure from Motion (SfM), and three-dimensional (3D) Digital Image Correlation (DIC) systems extremely appealing techniques for extracting structural deformations and geometry profiles. Therefore, optically based, non-contact measurement techniques may be used for assessing surface defects, rail and tie deflection profiles, and ballast condition. In this study, the design of two camera-based measurement systems is proposed for crossties-ballast condition assessment and track examination purposes. The first one consists of four pairs of cameras installed on the underside of a rail car to detect the induced deformation and displacement on the whole length of the track's cross tie using 3D DIC measurement techniques. The second consists of another set of cameras using SfM techniques for obtaining a 3D rendering of the infrastructure from a series of two-dimensional (2D) images to evaluate the state of the track qualitatively. The feasibility of the proposed optical systems is evaluated through extensive laboratory tests, demonstrating their ability to measure parameters of interest (e.g. crosstie's full-field displacement, vertical deflection, shape, etc.) for assessment and SHM of railroad track.

ASSESSMENT OF CABLE AGING USING CONDITION MONITORING TECHNIQUES

International Nuclear Information System (INIS)

GROVE, E.; LOFARO, R.; SOO, P.; VILLARAN, M.; HSU, F.

2000-01-01

Electric cables in nuclear power plants suffer degradation during service as a result of the thermal and radiation environments in which they are installed. Instrumentation and control cables are one type of cable that provide an important role in reactor safety. Should the polymeric cable insulation material become embrittled and cracked during service, or during a loss-of-coolant-accident (LOCA) and when steam and high radiation conditions are anticipated, failure could occur and prevent the cables from fulfilling their intended safety function(s). A research program is being conducted at Brookhaven National Laboratory to evaluate condition monitoring (CM) techniques for estimating the amount of cable degradation experienced during in-plant service. The objectives of this program are to assess the ability of the cables to perform under a simulated LOCA without losing their ability to function effectively, and to identify CM techniques which may be used to determine the effective lifetime of cables. The cable insulation materials tested include ethylene propylene rubber (EPR) and cross-linked polyethylene (XLPE). Accelerated aging (thermal and radiation) to the equivalent of 40 years of service was performed, followed by exposure to simulated LOCA conditions. The effectiveness of chemical, electrical, and mechanical condition monitoring techniques are being evaluated. Results indicate that several of these methods can detect changes in material parameters with increasing age. However, each has its limitations, and a combination of methods may provide an effective means for trending cable degradation in order to assess the remaining life of cables

Impact of structural aging on seismic risk assessment of reinforced concrete structures in nuclear power plants

International Nuclear Information System (INIS)

Ellingwood, B.; Song, J.

1996-03-01

The Structural Aging Program is addressing the potential for degradation of concrete structural components and systems in nuclear power plants over time due to aging and aggressive environmental stressors. Structures are passive under normal operating conditions but play a key role in mitigating design-basis events, particularly those arising from external challenges such as earthquakes, extreme winds, fires and floods. Structures are plant-specific and unique, often are difficult to inspect, and are virtually impossible to replace. The importance of structural failures in accident mitigation is amplified because such failures may lead to common-cause failures of other components. Structural condition assessment and service life prediction must focus on a few critical components and systems within the plant. Components and systems that are dominant contributors to risk and that require particular attention can be identified through the mathematical formalism of a probabilistic risk assessment, or PRA. To illustrate, the role of structural degradation due to aging on plant risk is examined through the framework of a Level 1 seismic PRA of a nuclear power plant. Plausible mechanisms of structural degradation are found to increase the core damage probability by approximately a factor of two

Assessing a new gene expression analysis technique for radiation biodosimetry applications

Energy Technology Data Exchange (ETDEWEB)

Manning, Grainne; Kabacik, Sylwia; Finnon, Paul; Paillier, Francois; Bouffler, Simon [Cancer Genetics and Cytogenetics, Biological Effects Department, Centre for Radiation, Chemical and Environmental Hazards, Health Protection Agency, Chilton, Didcot, Oxfordshire OX11 ORQ (United Kingdom); Badie, Christophe, E-mail: christophe.badie@hpa.org.uk [Cancer Genetics and Cytogenetics, Biological Effects Department, Centre for Radiation, Chemical and Environmental Hazards, Health Protection Agency, Chilton, Didcot, Oxfordshire OX11 ORQ (United Kingdom)

2011-09-15

The response to any radiation accident or incident involving actual or potential ionising radiation exposure requires accurate and rapid assessment of the doses received by individuals. The techniques available today for biodosimetry purposes are not fully adapted to rapid high-throughput measurements of exposures in large numbers of individuals. A recently emerging technique is based on gene expression analysis, as there are a number of genes which are radiation responsive in a dose-dependent manner. The present work aimed to assess a new technique which allows the detection of the level of expression of up to 800 genes without need of enzymatic reactions. In order to do so, human peripheral blood was exposed ex vivo to a range of x-ray doses from 5 mGy to 4 Gy of x-rays and the transcriptional expression of five radiation-responsive genes PHPT1, PUMA, CCNG1, DDB2 and MDM2 was studied by both the nCounter Digital Analyzer and Multiplex Quantitative Real-Time Polymerase Chain Reaction (MQRT-PCR) as the benchmark technology. Results from both techniques showed good correlation for all genes with R{sup 2} values ranging between 0.8160 and 0.9754. The reproducibility of the nCounter Digital Analyzer was also assessed in independent biological replicates and proved to be good. Although the slopes of the correlation of results obtained by the techniques suggest that MQRT-PCR is more sensitive than the nCounter Digital Analyzer, the nCounter Digital Analyzer provides sensitive and reliable data on modifications in gene expression in human blood exposed to radiation without enzymatic amplification of RNA prior to analysis.

A comparison of U.S. and European methods for accident scenario, identificaton, selection and quantification

International Nuclear Information System (INIS)

Cadwallader, L.C.; Djerassi, H.; Lampin, I.

1989-10-01

This paper presents a comparison of the varying methods used to identify and select accident-initiating events for safety analysis and probabilistic risk assessment (PRA). Initiating events are important in that they define the extent of a given safety analysis or PRA. Comprehensiveness in identification and selection of initiating events is necessary to ensure that a thorough analysis is being performed. While total completeness cannot ever be realized, inclusion of all safety significant events can be attained. The European approach to initiating event identification and selection arises from within a newly developed Safety Analysis methodology framework. This is a functional approach, with accident initiators based on events that will cause a system or facility loss of function. The US method divides accident initiators into two groups, internal accident initiators into two groups, internal and external events. Since traditional US PRA techniques are applied to fusion facilities, the recommended PRA-based approach is a review of historical safety documents coupled with a facility-level Master Logic Diagram. The US and European methods are described, and both are applied to a proposed International Thermonuclear Experiment Reactor (ITER) Magnet System in a sample problem. Contrasts in the US and European methods are discussed. Within their respective frameworks, each method can provide the comprehensiveness of safety-significant events needed for a thorough analysis. 4 refs., 8 figs., 11 tabs

Advanced Marketing Core Curriculum. Test Items and Assessment Techniques.

Science.gov (United States)

Smith, Clifton L.; And Others

This document contains duties and tasks, multiple-choice test items, and other assessment techniques for Missouri's advanced marketing core curriculum. The core curriculum begins with a list of 13 suggested textbook resources. Next, nine duties with their associated tasks are given. Under each task appears one or more citations to appropriate…

A Mechanistic Reliability Assessment of RVACS and Metal Fuel Inherent Reactivity Feedbacks

Energy Technology Data Exchange (ETDEWEB)

Grabaskas, David; Brunett, Acacia J.; Passerini, Stefano; Grelle, Austin

2017-09-24

GE Hitachi Nuclear Energy (GEH) and Argonne National Laboratory (Argonne) participated in a two year collaboration to modernize and update the probabilistic risk assessment (PRA) for the PRISM sodium fast reactor. At a high level, the primary outcome of the project was the development of a next-generation PRA that is intended to enable risk-informed prioritization of safety- and reliability-focused research and development. A central Argonne task during this project was a reliability assessment of passive safety systems, which included the Reactor Vessel Auxiliary Cooling System (RVACS) and the inherent reactivity feedbacks of the metal fuel core. Both systems were examined utilizing a methodology derived from the Reliability Method for Passive Safety Functions (RMPS), with an emphasis on developing success criteria based on mechanistic system modeling while also maintaining consistency with the Fuel Damage Categories (FDCs) of the mechanistic source term assessment. This paper provides an overview of the reliability analyses of both systems, including highlights of the FMEAs, the construction of best-estimate models, uncertain parameter screening and propagation, and the quantification of system failure probability. In particular, special focus is given to the methodologies to perform the analysis of uncertainty propagation and the determination of the likelihood of violating FDC limits. Additionally, important lessons learned are also reviewed, such as optimal sampling methodologies for the discovery of low likelihood failure events and strategies for the combined treatment of aleatory and epistemic uncertainties.

Application of Risk Assessment Tools in the Continuous Risk Management (CRM) Process

Science.gov (United States)

Ray, Paul S.

2002-01-01

Marshall Space Flight Center (MSFC) of the National Aeronautics and Space Administration (NASA) is currently implementing the Continuous Risk Management (CRM) Program developed by the Carnegie Mellon University and recommended by NASA as the Risk Management (RM) implementation approach. The four most frequently used risk assessment tools in the center are: (a) Failure Modes and Effects Analysis (FMEA), Hazard Analysis (HA), Fault Tree Analysis (FTA), and Probabilistic Risk Analysis (PRA). There are some guidelines for selecting the type of risk assessment tools during the project formulation phase of a project, but there is not enough guidance as to how to apply these tools in the Continuous Risk Management process (CRM). But the ways the safety and risk assessment tools are used make a significant difference in the effectiveness in the risk management function. Decisions regarding, what events are to be included in the analysis, to what level of details should the analysis be continued, make significant difference in the effectiveness of risk management program. Tools of risk analysis also depends on the phase of a project e.g. at the initial phase of a project, when not much data are available on hardware, standard FMEA cannot be applied; instead a functional FMEA may be appropriate. This study attempted to provide some directives to alleviate the difficulty in applying FTA, PRA, and FMEA in the CRM process. Hazard Analysis was not included in the scope of the study due to the short duration of the summer research project.

Assessment of Effective Prestressed Force of Nuclear Containment Building using SI Technique

Energy Technology Data Exchange (ETDEWEB)

Lee, H. P.; Jang, J. B.; Hwang, K. M.; Song, Y. C. [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

2010-10-15

Bonded tendons have been used in reactor buildings of heavy water reactors and the light water reactors of some nuclear power plants operating in Korea. The assessment of prestressed forces on those bonded tendons is becoming an important issue in assuring their continuous operation beyond their design life. In order to assess the effective prestressed force on the bonded tendon, indirect assessment techniques have been applying to the test beams which were manufactured on construction time. Therefore, this research mainly forced to establish the assessment methodology to measure directly the effective prestressed force on the bonded tendon of containment buildings using System Identification (SI) technique. To accomplish this purpose, simple SI method was proposed and adapted three dimensional finite element analysis of the 1:4 scale prestressed concrete containment vessel (PCCV) tested by Sandia National Laboratory in 2000

Assessment of biological leaf tissue using biospeckle laser imaging technique

Science.gov (United States)

Ansari, M. Z.; Mujeeb, A.; Nirala, A. K.

2018-06-01

We report on the application of an optical imaging technique, the biospeckle laser, as a potential tool to assess biological and medicinal plant leaves. The biospeckle laser technique is a non-invasive and non-destructive optical technique used to investigate biological objects. Just after their removal from plants, the torn leaves were used for biospeckle laser imaging. Quantitative evaluation of the biospeckle data using the inertia moment (IM) of the time history speckle pattern, showed that the IM can be utilized to provide a biospeckle signature to the plant leaves. It showed that leaves from different plants can have their own characteristic IM values. We further investigated the infected regions of the leaves that display a relatively lower biospeckle activity than the healthy tissue. It was easy to discriminate between the infected and healthy regions of the leaf tissue. The biospeckle technique can successfully be implemented as a potential tool for the taxonomy of quality leaves. Furthermore, the technique can help boost the quality of ayurvedic medicines.

Application of probabilistic risk assessment: Evaluating remedial alternatives at the Portland Harbor Superfund Site, Portland, Oregon, USA.

Science.gov (United States)

Ruffle, Betsy; Henderson, James; Murphy-Hagan, Clare; Kirkwood, Gemma; Wolf, Frederick; Edwards, Deborah A

2018-01-01

A probabilistic risk assessment (PRA) was performed to evaluate the range of potential baseline and postremedy health risks to fish consumers at the Portland Harbor Superfund Site (the "Site"). The analysis focused on risks of consuming fish resident to the Site containing polychlorinated biphenyls (PCBs), given that this exposure scenario and contaminant are the primary basis for US Environmental Protection Agency's (USEPA's) selected remedy per the January 2017 Record of Decision (ROD). The PRA used probability distributions fit to the same data sets used in the deterministic baseline human health risk assessment (BHHRA) as well as recent sediment and fish tissue data to evaluate the range and likelihood of current baseline cancer risks and noncancer hazards for anglers. Areas of elevated PCBs in sediment were identified on the basis of a geospatial evaluation of the surface sediment data, and the ranges of risks and hazards associated with pre- and postremedy conditions were calculated. The analysis showed that less active remediation (targeted to areas with the highest concentrations) compared to the remedial alternative selected by USEPA in the ROD can achieve USEPA's interim risk management benchmarks (cancer risk of 10 -4 and noncancer hazard index [HI] of 10) immediately postremediation for the vast majority of subsistence anglers that consume smallmouth bass (SMB) fillet tissue. In addition, the same targeted remedy achieves USEPA's long-term benchmarks (10 -5 and HI of 1) for the majority of recreational anglers. Additional sediment remediation would result in negligible additional risk reduction due to the influence of background. The PRA approach applied here provides a simple but adaptive framework for analysis of risks and remedial options focused on variability in exposures. It can be updated and refined with new data to evaluate and reduce uncertainty, improve understanding of the Site and target populations, and foster informed remedial decision

Boat sampling technique for assessment of ageing of components

International Nuclear Information System (INIS)

Kumar, Kundan; Shyam, T.V.; Kayal, J.N.; Rupani, B.B.

2006-01-01

Boat sampling technique (BST) is a surface sampling technique, which has been developed for obtaining, in-situ, metal samples from the surface of an operating component without affecting its operating service life. The BST is non-destructive in nature and the sample is obtained without plastic deformation or without thermal degradation of the parent material. The shape and size of the sample depends upon the shape of the cutter and the surface geometry of the parent material. Miniature test specimens are generated from the sample and the specimens are subjected to various tests, viz. Metallurgical Evaluation, Metallographic Evaluation, Micro-hardness Evaluation, sensitisation test, small punch test etc. to confirm the integrity and assessment of safe operating life of the component. This paper highlights design objective of boat sampling technique, description of sampling module, sampling cutter and its performance evaluation, cutting process, boat samples, operational sequence of sampling module, qualification of sampling module, qualification of sampling technique, qualification of scooped region of the parent material, sample retrieval system, inspection, testing and examination to be carried out on the boat samples and scooped region. (author)

Probabilistic risk assessment methodology for risk management and regulatory applications

International Nuclear Information System (INIS)

See Meng Wong; Kelly, D.L.; Riley, J.E.

1997-01-01

This paper discusses the development and potential applications of PRA methodology for risk management and regulatory applications in the U.S. nuclear industry. The new PRA methodology centers on the development of This paper discusses the time-dependent configuration risk profile for evaluating the effectiveness of operational risk management programs at U.S. nuclear power plants. Configuration-risk profiles have been used as risk-information tools for (1) a better understanding of the impact of daily operational activities on plant safety, and (2) proactive planning of operational activities to manage risk. Trial applications of the methodology were undertaken to demonstrate that configuration-risk profiles can be developed routinely, and can be useful for various industry and regulatory applications. Lessons learned include a better understanding of the issues and characteristics of PRA models available to industry, and identifying the attributes and pitfalls in the developement of risk profiles

Interest and limitations of projective techniques in the assessment of personality disorders.

Science.gov (United States)

Petot, J M

2000-06-01

Assessing personality disorders (PD) remains a difficult task because of persistent problems linked to concurrent validity of existing instruments, which are all structured interviews or self-report inventories. It has been advocated that indirect methods, projective techniques in particular, can strengthen PD assessment methods. The thematic apperception test (TAT) may be a significant adjuvant method of PD assessment.

Use of Formative Classroom Assessment Techniques in a Project Management Course

Science.gov (United States)

Purcell, Bernice M.

2014-01-01

Formative assessment is considered to be an evaluation technique that informs the instructor of the level of student learning, giving evidence when it may be necessary for the instructor to make a change in delivery based upon the results. Several theories of formative assessment exist, all which propound the importance of feedback to the student.…

Air Quality Assessment Using Interpolation Technique

Directory of Open Access Journals (Sweden)

Awkash Kumar

2016-07-01

Full Text Available Air pollution is increasing rapidly in almost all cities around the world due to increase in population. Mumbai city in India is one of the mega cities where air quality is deteriorating at a very rapid rate. Air quality monitoring stations have been installed in the city to regulate air pollution control strategies to reduce the air pollution level. In this paper, air quality assessment has been carried out over the sample region using interpolation techniques. The technique Inverse Distance Weighting (IDW of Geographical Information System (GIS has been used to perform interpolation with the help of concentration data on air quality at three locations of Mumbai for the year 2008. The classification was done for the spatial and temporal variation in air quality levels for Mumbai region. The seasonal and annual variations of air quality levels for SO2, NOx and SPM (Suspended Particulate Matter have been focused in this study. Results show that SPM concentration always exceeded the permissible limit of National Ambient Air Quality Standard. Also, seasonal trends of pollutant SPM was low in monsoon due rain fall. The finding of this study will help to formulate control strategies for rational management of air pollution and can be used for many other regions.

Assessing Dental Hygienists' Communication Techniques for Use with Low Oral Health Literacy Patients.

Science.gov (United States)

Flynn, Priscilla; Acharya, Amit; Schwei, Kelsey; VanWormer, Jeffrey; Skrzypcak, Kaitlyn

2016-06-01

This primary aim of this study was to assess communication techniques used with low oral health literacy patients by dental hygienists in rural Wisconsin dental clinics. A secondary aim was to determine the utility of the survey instrument used in this study. A mixed methods study consisting of a cross-sectional survey, immediately followed by focus groups, was conducted among dental hygienists in the Marshfield Clinic (Wisconsin) service area. The survey quantified the routine use of 18 communication techniques previously shown to be effective with low oral health literacy patients. Linear regression was used to analyze the association between routine use of each communication technique and several indicator variables, including geographic practice region, oral health literacy familiarity, communication skills training and demographic indicators. Qualitative analyses included code mapping to the 18 communication techniques identified in the survey, and generating new codes based on discussion content. On average, the 38 study participants routinely used 6.3 communication techniques. Dental hygienists who used an oral health literacy assessment tool reported using significantly more communication techniques compared to those who did not use an oral health literacy assessment tool. Focus group results differed from survey responses as few dental hygienists stated familiarity with the term "oral health literacy." Motivational interviewing techniques and using an integrated electronic medical-dental record were additional communication techniques identified as useful with low oral health literacy patients. Dental hygienists in this study routinely used approximately one-third of the communication techniques recommended for low oral health literacy patients supporting the need for training on this topic. Based on focus group results, the survey used in this study warrants modification and psychometric testing prior to further use. Copyright © 2016 The American Dental

The application of availability analysis to nuclear power plants

International Nuclear Information System (INIS)

Brooks, A.C.

1984-01-01

The use of probabilistic risk analysis (PRA) to assess the risks from nuclear power plants is now well established. Considerably less attention has been given so far to the use of availability analysis techniques. The economics of power generation are now such that with nuclear power currently supplying a substantial fraction of power in many countries, increasing attention is being paid to improving plant availability. This paper presents a technique for systematically identifying the areas in which measures to improve plant availability will be most effective. (author)

Reliability and criterion validity of an observation protocol for working technique assessments in cash register work.

Science.gov (United States)

Palm, Peter; Josephson, Malin; Mathiassen, Svend Erik; Kjellberg, Katarina

2016-06-01

We evaluated the intra- and inter-observer reliability and criterion validity of an observation protocol, developed in an iterative process involving practicing ergonomists, for assessment of working technique during cash register work for the purpose of preventing upper extremity symptoms. Two ergonomists independently assessed 17 15-min videos of cash register work on two occasions each, as a basis for examining reliability. Criterion validity was assessed by comparing these assessments with meticulous video-based analyses by researchers. Intra-observer reliability was acceptable (i.e. proportional agreement >0.7 and kappa >0.4) for 10/10 questions. Inter-observer reliability was acceptable for only 3/10 questions. An acceptable inter-observer reliability combined with an acceptable criterion validity was obtained only for one working technique aspect, 'Quality of movements'. Thus, major elements of the cashiers' working technique could not be assessed with an acceptable accuracy from short periods of observations by one observer, such as often desired by practitioners. Practitioner Summary: We examined an observation protocol for assessing working technique in cash register work. It was feasible in use, but inter-observer reliability and criterion validity were generally not acceptable when working technique aspects were assessed from short periods of work. We recommend the protocol to be used for educational purposes only.

Nuclear power plant risk assembly and decomposition for risk management

International Nuclear Information System (INIS)

Iden, D.C.

1985-01-01

The state-of-the-art method for analyzing the risk from nuclear power plants is probabilistic risk assessment (PRA). The intermediate results of a PRA are first assembled to quantify the risk from operating a nuclear power plant in the form of (1) core damage (or core melt) frequency, (2) plant damage state frequencies, (3) release category frequencies, and (4) the frequency of exceeding specific levels of offsite consequences. Once the overall PRA results have been quantified, the next step is to decompose those results into the individual contributors to each of the four forms of risk in some rank order. The way in which the PRA model is set up to assemble and decompose the plant risk determines the ease and usefulness of the PRA model as a risk management tool for evaluating perturbations to the PRA model. These perturbations can take the form of technical specification changes, hardware modifications, procedural changes, etc. The matrix formalism developed by Dr. Stan Kaplan for risk assembly and decomposition represents a significant breakthrough in making the PRA model an effective risk management tool. The key to understanding the matrix formalism and making it a useful tool for managing nuclear power plant risk is the structure of the PRA model. PRA risk model structure and decomposition of the risk results are discussed with the Seabrook PRA as an example

A new scintigraphic technique for assessment of right atrial function

International Nuclear Information System (INIS)

Marmor, A.; Biello, D.R.; Sampathkumaran, K.S.; Geltman, E.M.; Siegel, B.A.; Roberts, R.

1981-01-01

Gated radionuclide imaging with 99 mTc-labeled red blood cells was employed to assess changes in right atrial (RA) volume in 17 subjects. Studies were repeated within two hours in five of 12 normal subjects and at two and ten days after infarction in five patients. Time-activity curves were generated for the RA and right ventricle (RV) from variable regions of interest defined by automatic edge detection. The RA time-activity curve exhibited four phases: filling, early emptying, equilibration, and late emptying. Repeated measurements of RA early emptying rates correlated closely in normals (r . 0.93). Early RA emptying and RV filling rates corresponded closely (r . 0.89). In four patients with infarction, the RA emptying rate was slowed on the first study but had normalized by 10 days. In a patient with inferior infarction and suspected RV infarction, the RA emptying rate was depressed and remained so for 10 days. Thus, relative changes in right atrial volume can be assessed reproducibly by a radionuclide technique, which is relatively independent of geometric assumptions. This technique offers great promise as a method to assess atrial function in health and disease

Performing Probabilistic Risk Assessment Through RAVEN

Energy Technology Data Exchange (ETDEWEB)

A. Alfonsi; C. Rabiti; D. Mandelli; J. Cogliati; R. Kinoshita

2013-06-01

The Reactor Analysis and Virtual control ENviroment (RAVEN) code is a software tool that acts as the control logic driver and post-processing engine for the newly developed Thermal-Hydraulic code RELAP-7. RAVEN is now a multi-purpose Probabilistic Risk Assessment (PRA) software framework that allows dispatching different functionalities: Derive and actuate the control logic required to simulate the plant control system and operator actions (guided procedures), allowing on-line monitoring/controlling in the Phase Space Perform both Monte-Carlo sampling of random distributed events and Dynamic Event Tree based analysis Facilitate the input/output handling through a Graphical User Interface (GUI) and a post-processing data mining module

Choice and complexation of techniques and tools for assessment of NPP I and C systems safety

International Nuclear Information System (INIS)

Illiashenko, Oleg; Babeshko, Eugene

2011-01-01

There are a lot of techniques to analyze and assess reliability and safety of NPP Instrumentation and Control (I and C) systems (e.g. FMEA - Failure Modes and Effects Analysis and its modifications, FTA - Fault Tree Analysis, HAZOP - Hazard and Operability Analysis, RBD - Reliability Block Diagram, Markov Models, etc.) and quantity of tools based on these techniques is constantly increasing. Known ways of safety assessment, as well as problems of their choice and complexation are analyzed. Objective of the paper is the development of general 'technique of techniques choosing' and tool for support of such technique. The following criteria are used for analysis and comparison and their features are described: compliance to normative documents; experience of application in industry; methods used for assessment of system NPP I and C safety; tool architecture/framework; reporting; vendor support, etc. Comparative analysis results of existing T and T - Tools and Techniques for safety analysis are presented in matrix form ('Tools-Criterion') with example. Features of complexation of different safety assessment techniques (FMECA, FTA, RBD, Markov Models) are described. The proposed technique is implemented as special tool for decision-making. The proposed technique was used for development of RPC Radiy company standard CS 66. This guide contains requirements and procedures of FMECA analysis of developed and produced NPP I and C systems based on RADIY platform. (author)

Need to use probabilistic risk approach in performance assessment of waste disposal facilities

International Nuclear Information System (INIS)

Bonano, E.J.; Gallegos, D.P.

1991-01-01

Regulations governing the disposal of radioactive, hazardous, and/or mixed wastes will likely require, either directly or indirectly, that the performance of disposal facilities be assessed quantitatively. Such analyses, commonly called ''performance assessments,'' rely on the use of predictive models to arrive at a quantitative estimate of the potential impact of disposal on the environment and the safety and health of the public. It has been recognized that a suite of uncertainties affect the results of a performance assessment. These uncertainties are conventionally categorized as (1) uncertainty in the future state of the disposal system (facility and surrounding medium), (2) uncertainty in models (including conceptual models, mathematical models, and computer codes), and (3) uncertainty in data and parameters. Decisions regarding the suitability of a waste disposal facility must be made in light of these uncertainties. Hence, an approach is needed that would allow the explicit consideration of these uncertainties so that their impact on the estimated consequences of disposal can be evaluated. While most regulations for waste disposal do not prescribe the consideration of uncertainties, it is proposed that, even in such cases, a meaningful decision regarding the suitability of a waste disposal facility cannot be made without considering the impact of the attendant uncertainties. A probabilistic risk assessment (PRA) approach provides the formalism for considering the uncertainties and the technical basis that the decision makers can use in discharging their duties. A PRA methodology developed and demonstrated for the disposal of high-level radioactive waste provides a general framework for assessing the disposal of all types of wastes (radioactive, hazardous, and mixed). 15 refs., 1 fig., 1 tab

Project Administration Techniques for Successful Classroom Collaborative Writing.

Science.gov (United States)

Kryder, LeeAnne Giannone

1991-01-01

Focuses on the collaborative writing done for a large report or proposal over a period of several weeks or months in a business writing course. Discusses short-term writing projects and nonwriting tasks for project administration, meeting management, student/instructor conference, project planning and time estimates, and oral presentations. (PRA)

Reliability Assessment of Wind Farm Electrical System Based on a Probability Transfer Technique

Directory of Open Access Journals (Sweden)

Hejun Yang

2018-03-01

Full Text Available The electrical system of a wind farm has a significant influence on the wind farm reliability and electrical energy yield. The disconnect switch installed in an electrical system cannot only improve the operating flexibility, but also enhance the reliability for a wind farm. Therefore, this paper develops a probabilistic transfer technique for integrating the electrical topology structure, the isolation operation of disconnect switch, and stochastic failure of electrical equipment into the reliability assessment of wind farm electrical system. Firstly, as the traditional two-state reliability model of electrical equipment cannot consider the isolation operation, so the paper develops a three-state reliability model to replace the two-state model for incorporating the isolation operation. In addition, a proportion apportion technique is presented to evaluate the state probability. Secondly, this paper develops a probabilistic transfer technique based on the thoughts that through transfer the unreliability of electrical system to the energy transmission interruption of wind turbine generators (WTGs. Finally, some novel indices for describing the reliability of wind farm electrical system are designed, and the variance coefficient of the designed indices is used as a convergence criterion to determine the termination of the assessment process. The proposed technique is applied to the reliability assessment of a wind farm with the different topologies. The simulation results show that the proposed techniques are effective in practical applications.

The NUREG-1150 probabilistic risk assessment for the Sequoyah nuclear plant

International Nuclear Information System (INIS)

Gregory, J.J.; Breeding, R.J.; Higgins, S.J.; Shiver, A.W.; Helton, J.C.; Murfin, W.B.

1992-01-01

This paper summarizes the findings of the probabilistic risk assessment (PRA) for Unit 1 of the Sequoyah Nuclear Plant performed in support of NUREG-1150. The emphasis is on the 'back-end' analyses, the accident progression, source term, and consequence analyses, and the risk results obtained when the results of these analyses are combined with the accident frequency analysis. The results of this PRA indicate that the offsite risk from internal initiating events at Sequoyah are quite low with respect to the safety goals. The containment appears likely to withstand the loads that might be placed upon it if the reactor vessel fails. A good portion of the risk, in this analysis, comes from initiating events which bypass the containment. These events are estimated to have a relatively low frequency of occurrence, but their consequences are quite large. Other events that contribute to offsite risk involve early containment failures that occur during degradation of the core or near the time of vessel breach. Considerable uncertainty is associated with the risk estimates produced in this analysis. Offsite risk from external initiating events was not included in this analysis. (orig.)

An image processing technique for the radiographic assessment of vertebral derangements

Energy Technology Data Exchange (ETDEWEB)

Breen, A.C. (Anglo-European Coll. of Chiropractic, Bournemouth (UK)); Allen, R. (Southampton Univ. (UK). Dept. of Mechanical Engineering); Morris, A. (Odstock Hospital, Salisbury (UK). Dept. of Radiology)

1989-01-01

A technique for measuring inter-vertebral motion by the digitization and processing of intensifier images is described. The technique reduces the time and X-ray dosage currently required to make such assessments. The errors associated with computing kinematic indices at increments of coronal plane rotations in the lumbar spine have been calculated using a calibration model designed to produce a facsimile of in vivo conditions in terms of image quality and geometric distortion. (author).

Diablo Canyon internal events PRA [Probabilistic Risk Assessment] review: Methodology and findings

International Nuclear Information System (INIS)

Fitzpatrick, R.G.; Bozoki, G.; Sabek, M.

1990-01-01

The review of the Diablo Canyon Probabilistic Risk Assessment (DCRPA) incorporated some new and innovative approaches. These were necessitated by the unprecedented size, scope and level of detail of the DCRPA, which was submitted to the NRC for licensing purposes. This paper outlines the elements of the internal events portion of the review citing selected findings to illustrate the various approaches employed. The paper also provides a description of the extensive and comprehensive importance analysis applied by BNL to the DCRPA model. Importance calculations included: top event/function level; individual split fractions; pair importances between frontline-support and support-support systems; system importance by initiator; and others. The paper concludes with a brief discussion of the effectiveness of the applied methodology. 3 refs., 5 tabs

Use of probabilistic risk assessments to define areas of possible exemption from regulatory requirements

International Nuclear Information System (INIS)

Thompson, C.A.; Carlson, D.; Kolaczkowski, A.; LaChance, J.

1988-01-01

The Risk-Based Licensing Program (RBLP) was sponsored by the Department of Energy for the purpose of establishing and demonstrating an approach for identifying potential areas for exemption from current regulatory requirements in the licensing of nuclear power plants. Such an approach could assist in the improvement of the regulatory process for both current and future nuclear plant designs. Use of the methodology could result in streamlining the regulatory process by eliminating unnecessarily detailed reviews of portions of a plant design not important to risk. The RBLP methodology utilizes probabilistic risk assessments, (PRAs), which are required of all future applicants for nuclear power plant licenses. PRA results are used as a screening tool to determine the risk significance of various plant features which are correlated to the risk importance of regulations to identify potential areas for regulatory exemption. Additional consideration is then given to non-risk factors in the final determination of exemption candidates. The RBLP methodology was demonstrated using an existing PRA. The results of the demonstration are highlighted. 10 refs

Developing and Implementing an Assessment Technique to Measure Linked Concepts

Science.gov (United States)

Ye, Li; Oueini, Razanne; Lewis, Scott E.

2015-01-01

The links students make among chemistry content is considered essential for a robust, enduring understanding in multiple learning theories. This article describes the development and implementation of an assessment technique, termed a Measure of Linked Concepts, designed to inform instructors on students' understanding of linking content…

Review of the Diablo Canyon probabilistic risk assessment

International Nuclear Information System (INIS)

Bozoki, G.E.; Fitzpatrick, R.G.; Bohn, M.P.; Sabek, M.G.; Ravindra, M.K.; Johnson, J.J.

1994-08-01

This report details the review of the Diablo Canyon Probabilistic Risk Assessment (DCPRA). The study was performed under contract from the Probabilistic Risk Analysis Branch, Office of Nuclear Reactor Research, USNRC by Brookhaven National Laboratory. The DCPRA is a full scope Level I effort and although the review touched on all aspects of the PRA, the internal events and seismic events received the vast majority of the review effort. The report includes a number of independent systems analyses sensitivity studies, importance analyses as well as conclusions on the adequacy of the DCPRA for use in the Diablo Canyon Long Term Seismic Program

A historical overview of probabilistic risk assessment development and its use in the nuclear power industry: a tribute to the late Professor Norman Carl Rasmussen

International Nuclear Information System (INIS)

Keller, William; Modarres, Mohammad

2005-01-01

This paper reviews the historical development of the probabilistic risk assessment (PRA) methods and applications in the nuclear industry. A review of nuclear safety and regulatory developments in the early days of nuclear power in the United States has been presented. It is argued that due to technical difficulties for measuring and characterizing uncertainties and concerns over legal challenges, safety design and regulation of nuclear power plants has primarily relied upon conservative safety assessment methods derived based on a set of design and safety principles. Further, it is noted that the conservatism adopted in safety and design assessments has allowed the use of deterministic performance assessment methods. This approach worked successfully in the early years of nuclear power epoch as the reactor design proved to be safe enough. However, it has been observed that as the conservative approach to design and safety criteria proved arbitrary, and yielded inconsistencies in the degree to which different safety measures in nuclear power plants protect safety and public heath, the urge for a more consistent assessment of safety became apparent in the late 1960s. In the early 1970s, as a result of public and political pressures, then the US Atomic Energy Commission initiated a new look at the safety of the nuclear power plants through a comprehensive study called 'Reactor Safety Study' (WASH-1400, or 'Rasmussen Study'-after its charismatic study leader Professor Norman Rasmussen of MIT) to demonstrate safety of the nuclear power plants. Completed in October 1975, this landmark study introduced a novel probabilistic, systematic and holistic approach to the assessment of safety, which ultimately resulted in a sweeping paradigm shift in safety design and regulation of nuclear power in the United States in the turn of the Century. Technical issues of historic significance and concerns raised by the subsequent reviews of the Rasmussen Study have been discussed

An approach for using risk assessment in risk-informed decisions on plant-specific changes to the licensing basis

International Nuclear Information System (INIS)

Caruso, Mark A.; Cheok, Michael C.; Cunningham, Mark A.; Holahan, Gary M.; King, Thomas L.; Parry, Gareth W.; Ramey-Smith, Ann M.; Rubin, Mark P.; Thadani, Ashok C.

1999-01-01

This paper discusses an acceptable approach that the US Nuclear Regulatory Commission staff has proposed for using Probabilistic Risk Assessment in making decisions on changes to the licensing basis of a nuclear power plant. First, the overall philosophy of risk-informed decision-making, and the process framework are described. The philosophy is encapsulated in five principles, one of which states that, if the proposed change leads to an increase in core damage frequency or risk, the increases must be small and consistent with the intent of the Nuclear Regulatory Commission's Safety Goal Policy Statement. The second part of the paper discusses the use of PRA to demonstrate that this principle has been met. The discussion focuses on the acceptance guidelines, and on comparison of the PRA results with those guidelines. The difficulties that arise because of limitations in scope and analytical uncertainties are discussed and approaches to accommodate these difficulties in the decision-making are described

Multi-Hazard Advanced Seismic Probabilistic Risk Assessment Tools and Applications

International Nuclear Information System (INIS)

Coleman, Justin L.; Bolisetti, Chandu; Veeraraghavan, Swetha; Parisi, Carlo; Prescott, Steven R.; Gupta, Abhinav

2016-01-01

Design of nuclear power plant (NPP) facilities to resist natural hazards has been a part of the regulatory process from the beginning of the NPP industry in the United States (US), but has evolved substantially over time. The original set of approaches and methods was entirely deterministic in nature and focused on a traditional engineering margins-based approach. However, over time probabilistic and risk-informed approaches were also developed and implemented in US Nuclear Regulatory Commission (NRC) guidance and regulation. A defense-in-depth framework has also been incorporated into US regulatory guidance over time. As a result, today, the US regulatory framework incorporates deterministic and probabilistic approaches for a range of different applications and for a range of natural hazard considerations. This framework will continue to evolve as a result of improved knowledge and newly identified regulatory needs and objectives, most notably in response to the NRC activities developed in response to the 2011 Fukushima accident in Japan. Although the US regulatory framework has continued to evolve over time, the tools, methods and data available to the US nuclear industry to meet the changing requirements have not kept pace. Notably, there is significant room for improvement in the tools and methods available for external event probabilistic risk assessment (PRA), which is the principal assessment approach used in risk-informed regulations and risk-informed decision-making applied to natural hazard assessment and design. This is particularly true if PRA is applied to natural hazards other than seismic loading. Development of a new set of tools and methods that incorporate current knowledge, modern best practice, and state-of-the-art computational resources would lead to more reliable assessment of facility risk and risk insights (e.g., the SSCs and accident sequences that are most risk-significant), with less uncertainty and reduced conservatisms.

Multi-Hazard Advanced Seismic Probabilistic Risk Assessment Tools and Applications

Energy Technology Data Exchange (ETDEWEB)

Coleman, Justin L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bolisetti, Chandu [Idaho National Lab. (INL), Idaho Falls, ID (United States); Veeraraghavan, Swetha [Idaho National Lab. (INL), Idaho Falls, ID (United States); Parisi, Carlo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Prescott, Steven R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gupta, Abhinav [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-09-01

Design of nuclear power plant (NPP) facilities to resist natural hazards has been a part of the regulatory process from the beginning of the NPP industry in the United States (US), but has evolved substantially over time. The original set of approaches and methods was entirely deterministic in nature and focused on a traditional engineering margins-based approach. However, over time probabilistic and risk-informed approaches were also developed and implemented in US Nuclear Regulatory Commission (NRC) guidance and regulation. A defense-in-depth framework has also been incorporated into US regulatory guidance over time. As a result, today, the US regulatory framework incorporates deterministic and probabilistic approaches for a range of different applications and for a range of natural hazard considerations. This framework will continue to evolve as a result of improved knowledge and newly identified regulatory needs and objectives, most notably in response to the NRC activities developed in response to the 2011 Fukushima accident in Japan. Although the US regulatory framework has continued to evolve over time, the tools, methods and data available to the US nuclear industry to meet the changing requirements have not kept pace. Notably, there is significant room for improvement in the tools and methods available for external event probabilistic risk assessment (PRA), which is the principal assessment approach used in risk-informed regulations and risk-informed decision-making applied to natural hazard assessment and design. This is particularly true if PRA is applied to natural hazards other than seismic loading. Development of a new set of tools and methods that incorporate current knowledge, modern best practice, and state-of-the-art computational resources would lead to more reliable assessment of facility risk and risk insights (e.g., the SSCs and accident sequences that are most risk-significant), with less uncertainty and reduced conservatisms.

HTGR containment design options: an application of probabilistic risk assessment

International Nuclear Information System (INIS)

1977-08-01

Through the use of probabilistic risk assessment (PRA), it is possible to quantitatively evaluate the radiological risk associated with a given reactor design and to place such risk into perspective with alternative designs. The merits are discussed for several containment alternatives for the HTGR from the viewpoints of economics and licensability, as well as public risk. The quantification of cost savings and public risk indicates that presently acceptable public risk can be maintained and cost savings of $40 million can result from use of a vented confinement for the HTGR

Development of safety function assessment trees for pressurized heavy water reactor LP/SD operations

International Nuclear Information System (INIS)

Yang, Hui Chang; Chung, Chang Hyun; Kim, Ki Yong; Jee, Moon Hak; Sung, Chang Kyoung

2003-01-01

The objective of Configuration Risk Management Program(CRMP) is to maintain the safety level by assuring the defense-in-depth of nuclear power plant while the configurations are changed during plant operations, especially for the LP/SD. Such a safety purpose can be achieved by establishing the risk monitoring programs with both quantitative and qualitative features. Generally, the quantitative risk evaluation models, i.e., PRA models are used for the risk evaluation during full power operation, and the qualitative risk evaluation models such as safety function assessment trees are used. Through this study, safety function assessment trees were developed

Significance of earthquake risk in nuclear power plant probabilistic risk assessments

International Nuclear Information System (INIS)

Sues, R.H.; Amico, P.J.; Campbell, R.D.

1990-01-01

During the last eight years, approximately 25 utility-sponsored probabilistic risk assessments (PRAs) have been conducted for US nuclear reactors. Of these, ten have been published, seven of which have included complete seismic risk assessment. The results of the seven published PRAs are reviewed here in order to ascertain the significance of the risk due to earthquake initiating events. While PRA methodology has been in a state of development over the past seven years, and the results are subject to interpretation (as discussed in the paper), from the review conducted it is clear that earthquake-induced initiating events are important risk contributors. It is concluded that earthquake initiating events should not be dismissed, a priori, in any nuclear plant risk assessment. (orig.)

Nuclear medicine techniques in the assessment of alkaptonuria.

Science.gov (United States)

Vinjamuri, Sobhan; Ramesh, Chandakacharla N; Jarvis, Jonathan; Gallagher, Jim A; Ranganath, Lakshminarayana L

2011-10-01

Alkaptonuria is a rare autosomal recessive disorder due to a lack of the enzyme homogentisate dioxygenase, leading to ochronosis, a process of accumulation of a melanin-like polymer of homogentisic acid in cartilage and other collagenous structures. Patients develop severe osteoarthropathy that resembles osteoarthritis. Although the diagnosis of alkaptonuria is not particularly challenging in view of the blue-black discolouration of visible connective tissue and the presence of homogentisic acid in urine, the natural history of alkaptonuria remains poorly understood. Patients would benefit immensely from an objective assessment of their disease status and from a clearer understanding of the pathophysiology and associated physical changes. Isotope bone scans, which are commonly used to identify the extent of involvement of bones in cancerous processes, have also been increasingly used for characterizing the extent of arthropathy in conditions such as osteoarthritis and rheumatoid arthritis. Semiquantitative scores based on the extent of involvement of joints have been used to describe the involvement of large joints in the context of symptomatic treatment for osteoarthritis. A similar semiquantitative isotope bone scan score depending on the involvement of the number of large joints in patients with alkaptonuria can be formulated and validated in a suitable cohort of patients. Bone densitometry measurement using dual-energy X-ray absorptiometry scanning is an internationally accepted tool to assess the risk and extent of osteoporosis, and is increasingly used to assess the additional fracture risk in patients with arthropathy. We believe that, currently, nuclear medicine techniques can provide useful information, which can be incorporated into disease severity scores for alkaptonuria. Once the biological basis for alkaptonuria is better understood, it is feasible that nuclear medicine techniques of even greater sensitivity and specificity can be developed, thereby

Human Reliability Analysis: session summary

International Nuclear Information System (INIS)

Hall, R.E.

1985-01-01

The use of Human Reliability Analysis (HRA) to identify and resolve human factors issues has significantly increased over the past two years. Today, utilities, research institutions, consulting firms, and the regulatory agency have found a common application of HRA tools and Probabilistic Risk Assessment (PRA). The ''1985 IEEE Third Conference on Human Factors and Power Plants'' devoted three sessions to the discussion of these applications and a review of the insights so gained. This paper summarizes the three sessions and presents those common conclusions that were discussed during the meeting. The paper concludes that session participants supported the use of an adequately documented ''living PRA'' to address human factors issues in design and procedural changes, regulatory compliance, and training and that the techniques can produce cost effective qualitative results that are complementary to more classical human factors methods

Research on Methodology to Prioritize Critical Digital Assets based on Nuclear Risk Assessment

Energy Technology Data Exchange (ETDEWEB)

Kim, Wonjik; Kwon, Kookheui; Kim, Hyundoo [Korea Institute of Nuclear Nonproliferation and Control, Daejeon (Korea, Republic of)

2016-10-15

Digital systems are used in nuclear facilities to monitor and control various types of field devices, as well as to obtain and store vital information. Therefore, it is getting important for nuclear facilities to protect digital systems from cyber-attack in terms of safety operation and public health since cyber compromise of these systems could lead to unacceptable radiological consequences. Based on KINAC/RS-015 which is a cyber security regulatory standard, regulatory activities for cyber security at nuclear facilities generally focus on critical digital assets (CDAs) which are safety, security, and emergency preparedness related digital assets. Critical digital assets are estimated over 60% among all digital assets in a nuclear power plant. Therefore, it was required to prioritize critical digital assets to improve efficiency of regulation and implementation. In this paper, the research status on methodology development to prioritize critical digital assets based on nuclear risk assessment will be introduced. In this paper, to derive digital asset directly affect accident, PRA results (ET, FT, and minimal cut set) are analyzed. According to result of analysis, digital systems related to CD are derived ESF-CCS (safety-related component control system) and Process-CCS (non-safety-related component control system) as well as Engineered Safety Features Actuation System (ESFAS). These digital assets can be identified Vital Digital Asset (VDA). Hereafter, to develop general methodology which was identified VDA related to accident among CDAs, (1) method using result of minimal cut set in PRA model will be studied and (2) method quantifying result of Digital I and C PRA which is performed to reflect all digital cabinet related to system in FT will be studied.

Research on Methodology to Prioritize Critical Digital Assets based on Nuclear Risk Assessment